Establishment of control technology of the HTTR and future test plan

Honda, Yuki; Saito, Kenji; Tochio, Daisuke; Aono, Tetsuya; Hirato, Yoji; Kozawa, Takayuki; Nakagawa, Shigeaki

Journal of Nuclear Science and Technology, 51(11-12), p.1387 - 1397, 2014/11

https://doi.org/10.1080/00223131.2014.919241

The operational experiments of the HTTR would be useful for future high-temperature gas-cooled reactors (HTGRs). Main PID control constants of the HTTR are selected with reasonably damped characteristics and without undershoot or overshoot. For utilization the HTGR as a commercial reactor, it should be demonstrated that the HTGR system can supply stable heat to a heat utilization system for the long-term operation. The control characteristics in the long-term high-temperature operation are evaluated by the result of operation performed in 2010. In addition, from a viewpoint of HTGRs with heat utilization system, a future possibility of the experiments for heat utilization design is examined.

Performance-based improvement of the leakage rate test program for the reactor containment of HTTR; Adoption of revised test program containing "Type A, Type B and Type C tests"

Kondo, Masaaki; Kimishima, Satoru*; Emori, Koichi; Sekita, Kenji; Furusawa, Takayuki; Hayakawa, Masato; Kozawa, Takayuki; Aono, Tetsuya; Kuroha, Misao; Ouchi, Hiroshi

JAEA-Technology 2008-062, 46 Pages, 2008/10

JAEA-Technology-2008-062.pdf:11.62MB

The reactor containment of HTTR is tested to confirm leak-tight integrity of itself. "Type A test" has been conducted in accordance with the standard testing method in JEAC4203 since the preoperational verification of the containment was made. Type A tests are identified as basic one for measuring containment leakage rate, it costs much, however. Therefore, the test program for HTTR was revised to adopt an efficient and economical alternatives including "Type B and Type C tests". In JEAC4203-2004, following requirements are specified for adopting alternatives: upward trend of leakage rate by Type A test due to aging should not be recognized; criterion of combined leakage rate with Type B and Type C tests should be established; the criteria for Type A test and combined leakage rate test should be satisfied; correlation between the leakage rates by Type A test and combined leakage rate test should be recognized. Considering the performances of the tests, the policies of corresponding to the requirements were developed, which were accepted by the regulatory agency. This report presents an outline of the tests, identifies issues on the conventional test and summarizes the policies of corresponding to the requirements and of implementing the tests based on the revised program.

Leakage rate test for reactor containment vessel of HTTR

Kondo, Masaaki; Sekita, Kenji; Emori, Koichi; Sakaba, Nariaki; Kimishima, Satoru; Kuroha, Misao; Noji, Kiyoshi; Aono, Tetsuya; Hayakawa, Masato

JAEA-Testing 2006-002, 55 Pages, 2006/07

JAEA-Testing-2006-002.pdf:6.36MB

The leakage rate test for the reactor containment vessel of HTTR is conducted in accordance with the absolute pressure method provided in Japan Electric Association Code(JEAC4203). Although leakage test of a reactor containment vessel is, in general, performed in condition of reactor coolant pressure boundary to be opened in order to simulate an accident, the peculiar test method to HTTR which use the helium gas as reactor coolant has been established, in which the pressure boundary is closed to avoid the release of fission products into the environment of the reactor containment vessel. The system for measuring and calculating the data for evaluating the leakage rate for containment vessel of HTTR was developed followed by any modifications. Recently, the system has been improved for more accurate and reliable one with any useful functions including real time monitoring any conditions related to the test. In addition, the configuration of containment vessel boundary for the test and the calibration method for the detectors for measuring temperature in containment vessel have been modified by reflecting the revision of the Code mentioned above. This report describes the method, system configuration, and procedures for the leakage rate test for reactor containment vessel of HTTR.

Maintenance of fuel failure detector system of the HTTR

Noji, Kiyoshi; Kameyama, Yasuhiko; Emori, Koichi; Aono, Tetsuya

JAEA-Testing 2006-003, 47 Pages, 2006/06

JAEA-Testing-2006-003.pdf:7.43MB

The FFD (Fuel Failure Detection) System has been installed in the HTTR in order to detects the abnormal release of fission products from the fuel during the operations. The FFD system samples the primary coolant from the high-temperature plenum division of the reactor core divided into seven regions. The system detects short life fission product(FP) gases from each region. The damaged region can be specified by the FFD system. In the design, it was considered that the change in the sampling flow rate during operation was not necessary. However, it became clear that the measured value became unstable because of a fluctuation of the sampling flow rate due to change in the primary coolant pressure during operation. Moreover, it was difficult to change the sampling flow rate during operation. The sampling flow rate was controlled by manual valves located in the service area where the entry is limited during operation. Therefore, an improvement was carried out to control the sampling flow rate from the outside of the service area. The stable measured value was obtained by the improvement. Moreover, noise reduction, improvement of oil level gauge of compressors gives excellent operation of the FFD. This report summarizes the maintenance work of detectors (precipitator), equipment and improvement items of the system.

Oxygen Potentials of Pultonium and Uranium Mixed Oxide

Kato, Masato; Tamura, Tetsuya*; Aono, Shigenori

Proceedings of 11th Symposium on Thermodynamics of Nuclear Materials (STNM-11), P. 81, 2004/00

The oxygen potentials of (PuU)Owere measured by thermogravimetric equilibrium measurement in the range of the temperature from 800 to 1350 in Ar/H/HO or He/H/HO mixture gas flow. The experimental results of (PuU)Oare in good agree with the other works. The relationship between the partial oxygen pressure (PO) and X in MO was analized based on lattice defect theory. The oxygen potential of (PuU)Owas modeled by lattice defect theory using the data of the literature and this work. The resulting equation well reproduces the oxygen potential-temperature data for (PuU)O.

Oxidation rate of (UPu)O

Suzuki, Kiichi; Kato, Masato; Tamura, Tetsuya*; Uno, Hiroki*; Kashimura, Motoaki; Aono, Shigenori

no journal, , 

no abstracts in English

Characteristic confirmation test by using HTTR

Honda, Yuki; Aono, Tetsuya; Sekita, Kenji; Tochio, Daisuke; Takada, Shoji

no journal, , 

The Characteristic confirmation test has been demonstrating by using the High Temperature engineering Test Reactor (HTTR) to confirm the High Temperature Gas Reactor (HTGR) characteristics, which are safety characteristics and various heat utilization. The characteristic confirmation test by using the HTTR consists of the limit performance test and nuclear heat supply fluctuation test. The nuclear heat supply fluctuation testis planned to be carried out after restarting of the HTTR. Towards the realization of industrial utilization of a HTGR cogeneration system as an extension of a nuclear plant, it is important to ensure reactor safety in the case that thermal-load of the facility is fluctuated or lost. The HTTR is under long-term shutdown. However, two nuclear heat supply fluctuation tests were demonstrated without nuclear heating. This report shows outline, progress and results of characteristic confirmation test.