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Session Overview
Session
2.01: High Temperature Reactors-General & design, experiment
Time:
Wednesday, 18/Mar/2020:
11:30am - 1:00pm

Session Chair: Changho Lee, Argonne National Laboratory-ANL, United States of America
Session Chair: Seddon Atkinson, The University of Liverpool, United Kingdom
Location: R-1013

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Presentations

Economic Evaluation of the Proposed Advanced High Temperature Reactor Concept

Nthato Raboshaga1, Mmeli Fipaza2, Herbert Pelo2

1The South African Nuclear Energy Corporation; 2Eskom SOC Ltd.

The AHTR is the PBMR restart project, with unique plant configurations. The approach is to build the first machine, as a direct Helium Brayton cycle demonstration facility, on which the subsequent commercial plants will be based. However, the latter could be tailored to provide for power system flexibility, mobility, and other features that will be necessary to support grid demands of the future. The two design concepts are expected to exhibit variations primarily concerning versatility and costs.

This article is an effort to demonstrate the cost-implications and financial feasibility of the proposed AHTR concept configurations. The focus of the analyses is hinged on the appreciation of the combination of parameters including Levelized costs of electricity production, revenue, payback analysis, break-even analysis, plant capacity and availability, plant degradation and so forth. Since the AHTR is an evolutionary HTR, in the pre-conceptual design phase, the economic evaluation presented herein is a class four estimate. The research is qualitative, as it uses information from vendors in the industry.

The result shall be used in the project business case to determine the viability of this project. The commercial plant design is a multipurpose reactor with added capabilities to deal with the needs of the future such as heat application industries. Emphasis shall be on the Levelized Cost of electricity production, which allows for a fair comparison between electricity produced from various energy sources, it can also be regarded as the minimum cost at which electricity must be sold to break-even over the lifetime of the project. To ensure that the project is a success, cost of other energy production methods will be considered, as the plant competes with the latest technologies.



The main modules and features of HTR-STAC V2.0

Chuan Li1, Hongyu Chen1, Wenyi Wang1, Chao Fang2

1Tsinghua University; Collaborative Innovation Center of Advanced Nuclear Energy Technology; Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education; 2Lab for High Technology, Tsinghua University

High Temperature Reactor-Pebble bed Modules (HTR-PM) design by China is a typical HTGR (High Temperature Gas-Cooled Reactor) with Gen-IV features. 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-STAC (HTR-PM Source Term Analysis Code) has been developed by the Institute of Nuclear and New Energy Technology (INET), Tsinghua University. The HTR-STAC V1 consists of five units, including LOOP, NORMAL, ARCC, CARBON and TRUM. LOOP is used to study the primary circuit coolant radioactivity and NORMAL may be used as calculating the release of airborne radioactivity to the environment under Normal operating conditions of HTR-PM. The code ARCC is compiled based on the results given by LOOP and NORMAL, which focus on the radioactivity release under several design basic accidents, such as SGTR (Steam Generator Tube Rupture), LOCA (Loss of Coolant Accident) and the transient process. CARBON and TRUM are developed to calculate the productions of two significant radionuclides: C-14 and H-3. With respect to the previous Version 1, the new HTR-STAC V2.0 adds one more modules: FERRA (Fuel Element Release Rate Analysis code). FERRA is developed to study the fission products (FPs) release rate from the fuel elements. The results of FERRA can be used in LOOP to calculate the concentration of FPs in the primary circuit. Then, these above six modules of HTR-STAC could perform a more comprehensive source term analysis of HTR-PM, from the very beginning of FPs release to the radioactivity release under all situations.



Shielding Design and Dose Evaluation for HTR-PM Fuel Transport Pipelines by QAD-CGA Program

Wenqian Li1, Jianzhu Cao1, Chen Luo2, Sheng Fang1, Hong Li1, Shihai Ding3

1Tsinghua University; 2Beijing Institute of Metrology; 3China Nuclear Power Engineering CO., LTD

The spherical fuel elements are adopted in the high-temperature gas-cooled reactor pebble-module (HTR-PM). The fuel element will be discharged continuously from the reactor core and transported into the fuel transport pipelines during the reactor operation. The dose outside of the pipeline should be considered. It is a dynamic source term from the transport of the spherical fuel elements. Also, the length of the pipelines varies from place to place. In this work, the QAD-CGA program was adopted for the dose calculation, and the shielding design of the pipeline was made upon the calculation results. During the calculation, it is assumed that a spherical fuel element stays in different positions of the pipelines in turn, and the dose contributions of the fuel element to points of interest were calculated. Finally, the dose contributions at different dwell positions were integrated. It is assumed that the fuel elements transport speed is 5 m/s and 6,000 balls will be transported per day. Two types of fuel transport pipelines were considered: the spent fuel element transport pipelines and different burn-up fuel element transport pipelines. Correspondingly, two source terms were adopted: the spent fuel source term and the average burn-up fuel source term. Doses at different points of interest were calculated for with no shielding scenario and with lead shielding of different thicknesses scenario. To evaluate the shielding effect, the dose limit of the orange radiation zone of HTR-PM and the radiation damage thresholds from the NCRP report No.51 were both adopted. The calculated results show that for pipelines that transport the spent fuel, a 4cm lead shielding will be enough. And for pipelines that transport fuel elements with different burn-up, a 5cm lead shielding will be added. The method and results can provide valuable reference for other work.



The Development of a Control Strategy of a Generic Micro High Temperature Reactor

Seddon Atkinson1, Dzianis Litskevich1, Anna Detkina1, Bruno Merk1, Takeshi Aoki2

1The University of Liverpool; 2Japanese Atomic Energy Agency

To resolve the global energy crisis, the world is turning to privately affordable, versatile SMRs which has seen the rise in HTR technology. The control method is implemented on the conceptual design of the U-Battery, where the market potential is heavily based on remote locations. This provides issues with regards core design as the high fissile loading content required for long term operation becomes challenging to control.

Through my PhD we have managed to provide high fidelity models of the U-Battery, which highlighted this issue. The application of large amounts of fixed burnable poisons causes a high penalty on the maximum achievable full power days of the fuel cycle and this requires innovative ways to overcome this issue. We determined that one solution would be a moveable moderator (MM), which takes advantage of the thermal HTR nature of the core by removing a significant volume of moderating material.

The MM concept was proven to breed more plutonium and therefore enhance the full power days within the core. In addition to this, the MM reduced the initial keff by 3000 pcm. This paper investigates how the MM can be used in conjunction with standard reactivity control methods over the fuel cycle.

More recent work covers how the MM effects the safety performance during transients. HTRs passive safety features benefit from the high volumetric heat capacity of graphite to maintain a lower maximum temperature during a loss of coolant accident. This investigation determines the negative effects of removing graphite from the core and how this effects the maximum fuel temperatures during accident conditions.



 
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