The TINTE Modular Code System for Computational Simulation of Transient Processes in the Primary Circuit of a Pebble-Bed High-Temperature Gas-Cooled Reactor

Water-ingress accident, caused by the steam generator heating tube rupture of a high temperature gas-cooled reactor, will introduce a positive reactivity to lead the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Increase of the primary circuit pressure may result in the opening of the safety valve, which will cause the release of radioactive isotopes and flammable water gas. The analysis of such an important and particular accident is significant for verifying the inherent safety characteristics of the pebble-bed modular high temperature gas-cooled reactor. Based on the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the design basis accident of double-ended guillotine break of a heating tube has been analyzed by using TINTE, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature and primary loop pressure, the graphite corrosion inventory, the water gas releasing amount, as well as the natural convection influence under the condition of the failure of the blower flaps shut down, have been studied in detail. The calculation result of the design basis accident indicates that, the maximal possible water ingress amount is less than 600 kg and the maximal fuel temperature keeps far below the design limitation of 1620°C. The result also shows that the slight amount of graphite corrosion will not damage the reactor structure and the fuel element, and there is no potential explosive risk caused by the opening of the safety valve.

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The R&D of HTR-STAC Program Package: Source Term Analysis Codes for Pebble-Bed High-Temperature Gas-Cooled Reactor

Science and Technology of Nuclear Installations ◽

10.1155/2018/7389121 ◽

2018 ◽

Vol 2018 ◽

pp. 1-9 ◽

Cited By ~ 1

Author(s):

Hongyu Chen ◽

Chuan Li ◽

Haoyu Xing ◽

Chao Fang

Keyword(s):

High Temperature ◽

Nuclear Reactor ◽

Source Term ◽

Safety Analysis ◽

Operating Conditions ◽

Primary Circuit ◽

Pebble Bed ◽

Water Reactors ◽

Term Analysis ◽

High Temperature Gas

Source term analysis is important in the design and safety analysis of advanced nuclear reactor and also provides a radiation safety analysis basis for Modular High-Temperature Gas-Cooled Reactor (HTR). High-Temperature Gas-Cooled Reactor-Pebble-bed Modules (HTR-PM) design by China is a typical Gen-IV and due to different safety concepts and systems, the implements of source term analysis in light water reactors are not entirely applicable to HTR-PM. To solve this problem, HTR-PM Source Term Analysis Code (HTR-STAC) has been developed and related V&V has been finished. HTR-STAC consists of five units, including LOOP (Primary Circuit Source Term Analysis Code), NORMAL (Normal Condition Airborne Source Term Analysis Code), ARCC (Accident Release Category Calculation code), CARBON (C-14 Source Term Analysis Code), and TRUM (Tritium Source Term Analysis Code). LOOP and NORMAL may be used as calculating primary circuit coolant radioactivity and the release of airborne radioactivity to the environment under normal operating conditions of HTR-PM, respectively. The code ARCC composed of several source term analysis programs in the different typical accidents scenario, including SGTR (Steam Generator Tube Rupture), LOCA (Loss of Coolant Accident), and the Transient Process, is compiled based on the results given by LOOP and NORMAL. CARBON and TRUM are developed to calculate the productions of C-14 and H-3 through a different mechanism. Furthermore, the V&V has been performed and show some positive results.

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Study on on-line temperature measurement technology for core of pebble bed high temperature gas-cooled reactor

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2020.110944 ◽

2021 ◽

Vol 371 ◽

pp. 110944

Author(s):

Youjie Zhang ◽

Xiang Fang ◽

Tao Ma ◽

Bing Xia ◽

Chuan Li

Keyword(s):

High Temperature ◽

Temperature Measurement ◽

Measurement Technology ◽

Pebble Bed ◽

On Line ◽

Line Temperature ◽

High Temperature Gas

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POSCA: A computer code for fission product plateout and circulating coolant activities within the primary circuit of a high temperature gas-cooled reactor

Nuclear Engineering and Technology ◽

10.1016/j.net.2020.02.006 ◽

2020 ◽

Vol 52 (9) ◽

pp. 1974-1982

Author(s):

Nam-il Tak ◽

Jeong-Hun Lee ◽

Sung Nam Lee ◽

Chang Keun Jo

Keyword(s):

High Temperature ◽

Fission Product ◽

Computer Code ◽

Primary Circuit ◽

High Temperature Gas

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Design of the Online Gross γ Monitoring Instrument at the Exit of the Helium Purification System in HTR-PM

Science and Technology of Nuclear Installations ◽

10.1155/2018/5808352 ◽

2018 ◽

Vol 2018 ◽

pp. 1-12 ◽

Cited By ~ 1

Author(s):

Mengqi Lou ◽

Liguo Zhang ◽

Feng Xie ◽

Jianzhu Cao ◽

Jiejuan Tong ◽

...

Keyword(s):

High Temperature ◽

Detection Efficiency ◽

Primary Circuit ◽

Source Terms ◽

Primary Loop ◽

Monitoring Instrument ◽

Purification System ◽

Design Values ◽

Under Construction ◽

High Temperature Gas

After the successful construction and operation experience of the 10 MW high-temperature gas-cooled reactor (HTR-10), a high-temperature gas-cooled pebble-bed modular (HTR-PM) demonstration plant is under construction in Shidao Bay, Rongcheng City, Shandong province, China. An online gross γ monitoring instrument has been designed and placed at the exit of the helium purification system (HPS) of HTR-PM and is used to detect the activity concentration in the primary circuit after purification. The source terms in the primary loop of HTR-PM and the helium purification process were described. The detailed configuration of the gross γ monitoring instrument was presented in detail. The Monte Carlo method was used to simulate the detection efficiency of the monitoring system. Since the actual source terms in the primary loop of HTR-PM may be different than the current design values, a sensitivity analysis of the detection efficiency was implemented based on different relative proportions of the nuclides. The accuracy and resolution of the NaI(Tl) detector were discussed as well.

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