Core Design Study of Small-Sized High Temperature Reactor for Electricity Generation

2011 ◽  
Author(s):  
Minoru Goto ◽  
Satoshi Shimakawa ◽  
Atsuhiko Terada ◽  
Taiju Shibata ◽  
Yukio Tachibana ◽  
...  
Keyword(s):  
High Temperature ◽  
Reactor Power ◽  
Light Water Reactors ◽  
Core Size ◽  
Fuel Loading ◽  
The Core ◽  
Fuel Burn ◽  
Water Reactors ◽  
Test Reactor

A High Temperature Gas-cooled Reactor (HTGR) has several features different from conventional light water reactors such as inherent safety characteristics, high thermal efficiency and high economy. On the other hand, one of disadvantages of the HTGR with a prismatic core is to require rather long-term and expensive refueling, resulting in relatively long maintenance period and high cost. To solve the disadvantage, the present study challenges the core design of a small-sized reactor for long refueling interval by increasing core size, fuel loading and fuel burn up compared with the High Temperature engineering Test Reactor (HTTR). The preliminary burn-up calculation suggested that approximately 6 years of long refueling interval was found to be reasonably achieved. A refueling interval longer than 6 years may be possible by decreasing further power density, subsequently larger core size with operational reactor power of 120MWt, but this idea was not taken by the requirement of the reactor that the core size shall be accommodated reasonably in the core with double size of the HTTR at maximum.

Approaches to Process Monitoring in Small Modular Reactors

2014 ◽  
Author(s):  
B. R. Upadhyaya ◽  
C. Mehta ◽  
V. B. Lollar ◽  
J. W. Hines ◽  
D. de Wet
Keyword(s):  
Critical Parameters ◽  
Light Water Reactors ◽  
Term Operation ◽  
Severe Accidents ◽  
Small Modular Reactors ◽  
Flow Test ◽  
Test Loop ◽  
Pressure Water ◽  
Water Reactors

One of the advantages of small modular reactors (SMRs) is their possible deployment in remote locations and continued long-term operation with minimum downtime. In order to achieve this operational goal, the SMRs may require remote and continuous monitoring of process parameters. This feature is also important in monitoring critical parameters during severe accidents and for post-accident recovery. Small integral light water reactors have in-vessel space constraints and many of the traditional instrumentation are not practical in actual implementation. In order to resolve this issue, experiments were carried out on a flow test loop to characterize the relationship among process variables (flow rate, pressure, water level) and pump motor signatures. The paper presents the findings of this research with implications in relating electrical signatures to pump parameters.

Advanced Structural Materials and Cladding

MRS Bulletin ◽  
2009 ◽  
Vol 34 (1) ◽  
pp. 20-27 ◽  
Author(s):  
T. Allen ◽  
H. Burlet ◽  
R.K. Nanstad ◽  
M. Samaras ◽  
S. Ukai
Keyword(s):  
High Temperature ◽  
Structural Materials ◽  
Ceramic Composites ◽  
Pressure Vessels ◽  
Austenitic Steels ◽  
High Temperature Strength ◽  
Light Water Reactors ◽  
Modeling Tools ◽  
Light Water ◽  
Water Reactors

AbstractAdvanced nuclear energy systems, both fission- and fusion-based, aim to operate at higher temperatures and greater radiation exposure levels than experienced in current light water reactors. Additionally, they are envisioned to operate in coolants such as helium and sodium that allow for higher operating temperatures. Because of these unique environments, different requirements and challenges are presented for both structural materials and fuel cladding. For core and cladding applications in intermediate-temperature reactors (400–650°C), the primary candidates are 9–12Cr ferritic–martensitic steels (where the numbers represent the weight percentage of Cr in the material, i.e., 9–12 wt%) and advanced austenitic steels, adapted to maximize high-temperature strength without compromising lower temperature toughness. For very high temperature reactors (>650°C), strength and oxidation resistance are more critical. In such conditions, high-temperature metals as well as ceramics and ceramic composites are candidates. For all advanced systems operating at high pressures, performance of the pressure boundary materials (i.e., those components responsible for containing the high-pressure liquids or gases that cool the reactor) is critical to reactor safety. For some reactors, pressure vessels are anticipated to be significantly larger and thicker than those used in light water reactors. The properties through the entire thickness of these components, including the effects of radiation damage as a function of damage rate, are important. For all of these advanced systems, optimizing the microstructures of candidate materials will allow for improved radiation and high-temperature performance in nuclear applications, and advanced modeling tools provide a basis for developing optimized microstructures.

Thermal Evaluations for Next Generation Nuclear Plant Fuel Testing

Volume 4 ◽  
2004 ◽  
Author(s):  
Richard G. Ambrosek ◽  
Debbie J. Utterbeck ◽  
Brandon Miller
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Nuclear Plant ◽  
Light Water Reactors ◽  
Next Generation ◽  
Preliminary Information ◽  
Advanced Fuel ◽  
Water Reactors ◽  
Test Reactor ◽  
Test Fuel

The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility.

Life Beyond 60 Years: The Importance of Sustaining the Current Fleet of Light Water Reactors in the U.S. and the R&D Needs

2010 ◽  
Author(s):  
Ronaldo Szilard ◽  
Hongbin Zhang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Superior Performance ◽  
Light Water Reactors ◽  
Economic Operation ◽  
Significant Interest ◽  
Water Reactors ◽  
The U.S

The current fleet of 104 nuclear power plants in the U.S. began their operation with 40 years operating licenses. About half of these plants have their licenses renewed to 60 years and most of the remaining plants are anticipated to pursue license extension to 60 years. With the superior performance of the current fleet and formidable costs of building new nuclear power plants, there has been significant interest to extend the lifetime of the current fleet even further from 60 years to 80 years. This paper addresses some of the key long term technical challenges and identifies R&D needs related to the long term safe and economic operation of the current fleet.

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.

Sign in / Sign up

Export Citation Format

Share Document