Principle Design of Graphite Components for HTTR and R&D on Nuclear Graphite for HTGR in JAEA

2016 ◽  
Vol 697 ◽  
pp. 797-806
Author(s):  
Junya Sumita ◽  
Taiju Shibata ◽  
Tatsuo Iyoku ◽  
Masahiro Ishihara ◽  
Tetsuo Nishihara
Keyword(s):  
High Temperature ◽  
Structural Design ◽  
Generation Process ◽  
Design Code ◽  
Design Data ◽  
Inspection Method ◽  
The Core ◽  
Nuclear Graphite ◽  
Core Components ◽  
High Temperature Reactor

Nuclear energy is one of the most promising energy sources to satisfy energy security, environmental protection, and efficient supply. The High Temperature Gas-cooled Reactor (HTGR) has attractive inherent safety features and it can be used as many kinds of heat applications such as hydrogen production, electricity generation, process heat supply, district heating and desalination. Many countries, especially developing countries, show their interests in HTGR. Graphite materials are used for the core components of the HTGR. IG-110 graphite, fine-grained isotropic graphite, with high strength and high oxidation resistance is used in the High temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) and High Temperature Reactor-Pebble-bed Modules (HTR-PM) in China. IG-110 graphite is a major candidate for the core graphite components of the Very High Temperature Reactor (VHTR) which is one of HTGRs and one of the most promising candidates as the Generation-IV nuclear reactor systems. This paper describes design of core components of HTTR and R&D on nuclear graphite for HTGR. JAEA established the graphite structural design code and inspection standard of graphite to construct the HTTR. JAEA developed an in-service inspection method and a draft graphite structural design code for future HTGR on the basis of the HTTR technologies. Moreover, JAEA are now developing the design data base of IG-110 graphite and IG-430 graphite including irradiation data for HTGR.

Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs

2012 ◽  
Author(s):  
Kuniyoshi Takamatsu ◽  
Kazuhiro Sawa
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Control ◽  
Reactor Power ◽  
Reactor Kinetics ◽  
The Core ◽  
Power Control System ◽  
Stop Test ◽  
High Temperature Reactor ◽  
Forced Cooling

The High-Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas-cooled Reactor (HTGR) with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of JAEA. At present, test studies are being conducted using the HTTR to improve HTGR technologies in collaboration with domestic industries that also contribute to foreign projects for the acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are currently under development using data obtained with the HTTR, which include reactor kinetics, thermal hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). In this study, a three gas circulator trip test and a vessel cooling system (VCS) stop test were performed as a loss of forced cooling (LOFC) test to demonstrate the inherent safety features of HTGR. The VCS stop test involved stopping the VCS located outside the reactor pressure vessel to remove the residual heat of the reactor core as soon as the three gas circulators are tripped. All three gas circulators were tripped at 9, 24 and 30 MW. The primary coolant flow rate was reduced from the rated 45 t/h to 0 t/h. Control rods (CRs) were not inserted into the core and the reactor power control system was not operational. In fact, the three gas circulator tripping test at 9 MW has already been performed in a previous study. However, the results cannot be disclosed to the public because of a confidentiality agreement. Therefore, we cannot refer to the difference between the analytical and test results. We determined that the reactor power immediately decreases to the decay heat level owing to the negative reactivity feedback effect of the core, although the reactor shutdown system was not operational. Moreover, the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Core dynamics analysis of the LOFC test for the HTTR was performed. The relationship among the reactivities (namely, Doppler, moderator temperature, and xenon reactivities) affecting recriticality time and reactor peak power level as well as total reactivity was addressed. Furthermore, the analytical results for a reactor transient of hundred hours are presented. Based on the results, emergency operating procedures can be developed for the case of a loss of coolant accident in HTGR when the CRs are not inserted into the core and the reactor power control system is not operational. The analytical results will be used in the design and construction of the Kazakhstan High-Temperature Reactor and the realization of commercial Very High-Temperature Reactor systems.

Developments of metallic materials and a high-temperature structural design code for the HTTR

1991 ◽  
Vol 132 (1) ◽  
pp. 1-11 ◽  
Author(s):  
Kazuhiko Hada ◽  
Isoharu Nishiguchi ◽  
Yasushi Muto ◽  
Hirokazu Tsuji
Keyword(s):  
High Temperature ◽  
Structural Design ◽  
Metallic Materials ◽  
Design Code

Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Kaichao Sun ◽  
Lin-Wen Hu ◽  
Charles Forsberg
Keyword(s):  
High Temperature ◽  
Peak Power ◽  
Fuel Cycle ◽  
Thermal Power ◽  
Combined Cycle ◽  
Fluoride Salt ◽  
Control Rod ◽  
The Core ◽  
Liquid Salt ◽  
High Temperature Reactor

The fluoride-salt-cooled high-temperature reactor (FHR) is a new reactor concept, which combines low-pressure liquid salt coolant and high-temperature tristructural isotropic (TRISO) particle fuel. The refractory TRISO particle coating system and the dispersion in graphite matrix enhance safeguards (nuclear proliferation resistance) and security. Compared to the conventional high-temperature reactor (HTR) cooled by helium gas, the liquid salt system features significantly lower pressure, larger volumetric heat capacity, and higher thermal conductivity. The salt coolant enables coupling to a nuclear air-Brayton combined cycle (NACC) that provides base-load and peak-power capabilities. Added peak power is produced using jet fuel or locally produced hydrogen. The FHR is, therefore, considered as an ideal candidate for the transportable reactor concept to provide power to remote sites. In this context, a 20-MW (thermal power) compact core aiming at an 18-month once-through fuel cycle is currently under design at Massachusetts Institute of Technology (MIT). One of the key challenges of the core design is to minimize the reactivity swing induced by fuel depletion, since excessive reactivity will increase the complexity in control rod design and also result in criticality risk during the transportation process. In this study, burnable poison particles (BPPs) made of B4C with natural boron (i.e., 20% B10 content) are adopted as the key measure for fuel cycle optimization. It was found that the overall inventory and the individual size of BPPs are the two most important parameters that determine the evolution path of the multiplication factor over time. The packing fraction (PF) in the fuel compact and the height of active zone are of secondary importance. The neutronic effect of Li6 depletion was also quantified. The 18-month once-through fuel cycle is optimized, and the depletion reactivity swing is reduced to 1 beta. The reactivity control system, which consists of six control rods and 12 safety rods, has been implemented in the proposed FHR core configuration. It fully satisfies the design goal of limiting the maximum reactivity worth for single control rod ejection within 0.8 beta and ensuring shutdown margin with the most valuable safety rod fully withdrawn. The core power distribution including the control rod’s effect is also demonstrated in this paper.

Aerosol formation by graphite corrosion in case of water and air ingress to the core of a high-temperature reactor

Energy ◽  
1991 ◽  
Vol 16 (1-2) ◽  
pp. 491-499
Author(s):  
K. Kugeler ◽  
Ch. Epping ◽  
P. Schmidtlein ◽  
P. Schreiner
Keyword(s):  
High Temperature ◽  
Aerosol Formation ◽  
The Core ◽  
High Temperature Reactor ◽  
Air Ingress

CFD Analysis on the Influence of Coolant Distribution Blocks on the Core Hot Spot Temperature in a Prismatic Very High Temperature Reactor

2011 ◽  
Author(s):  
Min-Hwan Kim ◽  
Nam-il Tak ◽  
Jae Man Noh ◽  
Goon-Cherl Park
Keyword(s):  
High Temperature ◽  
Hot Spot ◽  
Maximum Velocity ◽  
Unit Cell ◽  
Normal Operation ◽  
The Core ◽  
Very High Temperature Reactor ◽  
Velocity Deviation ◽  
High Temperature Reactor ◽  
Very High

Two design options of core distribution block (CDB) for a cooled-vessel design in the Very High Temperature Reactor (VHTR) were developed and the influence on the core hot spot was investigated by the commercial computational fluid dynamics (CFD) code, CFX-11. Isothermal CFD analyses were performed to estimate the coolant flow variation at the inlet of the coolant channel. The results predicted about 5% of the maximum velocity deviation when applying the core pressure drop of NHDD PMR200. A unit-cell CFD model was used to access the effect of the velocity deviation on the core hot spot. The unit-cell analyses were carried out for the velocity deviation of 0%, 5%, and 10%. Not only a constant power but also a local maximum power profile was considered. According to the results, the maximum fuel temperature was increased by about 30°C for the velocity deviation of 10% but still below the normal operation limit of 1250°C.

scholarly journals Report on the ASME design code development effort for Composite Core Components for High Temperature Reactors

2018 ◽  
Author(s):  
Josina Wilna Geringer ◽  
Mark Mitchell ◽  
Timothy D. Burchell ◽  
Andrew A. Wereszczak ◽  
Yutai Kato
Keyword(s):  
High Temperature ◽  
Development Effort ◽  
Design Code ◽  
Core Components ◽  
High Temperature Reactors ◽  
Code Development
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