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JAEA Reports

Improvement in oil seal performance of gas compressor in HTTR, 2

Nemoto, Takahiro; Kaneshiro, Noriyuki*; Sekita, Kenji; Furusawa, Takayuki; Kuroha, Misao; Kawakami, Satoru; Kondo, Masaaki

JAEA-Technology 2015-006, 36 Pages, 2015/03

JAEA-Technology-2015-006.pdf:16.77MB

The High-Temperature engineering Test Reactor (HTTR) has been developed for establishing and upgrading the technical basis of HTGR.HTTR facilities have their structures, systems and a lot of components including reciprocating gas compressors, commonly used to extract and/or discharge reactor coolant helium gas contained in primary/secondary coolant systems. From the fact of the operational experiences of these compressors, seal-oil leakage has been frequently observed, although rod-seal mechanisms with complicated structures are equipped and improved for preventing coolant helium gas. So, we tried to change the rod-seal materials which might be a primary reason of frequent seal-oil leakage, that resulted in decreasing a mass and frequently of seal-oil leakage. It is confirmed that it is important to select adequate materials of rod seal for sliding speed of the piston of the compressor to prevent seal-oil leakage. Additionally, the procedure to estimate seal-oil leakage for each compressor is discussed. This report describes the results of investigation for improvement on seal-oil leak tightness of the compressors in HTTR facilities.

JAEA Reports

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.

JAEA Reports

Water chemistry control in HTTR

Sekita, Kenji; Furusawa, Takayuki; Emori, Koichi; Ishii, Taro*; Kuroha, Misao; Hayakawa, Masato; Ouchi, Hiroshi

JAEA-Technology 2008-057, 45 Pages, 2008/08

JAEA-Technology-2008-057.pdf:12.0MB

A carbon steel used is used for the main material for the components and pipings of the pressurized water cooling system etc. that are the reactor cooling system of the HTTR. Water quality is managed by using the hydrazine in the coolant of the water cooling system to prevent corrosion of the components and deoxidize the coolant. Also, regular analysis is carried out for the confirmation of the water quality. The following results were obtained through the water quality analysis. (1) In the pressurized water cooling system, the coolant temperature rises higher due to the heat removal of the primary coolant. So, the ammonia was formed in the thermal decomposition of the hydrazine. The electric conductivity increased, while the concentration of the hydrazine decreased, there was no problem as the plan it. (2) Thermal decomposition of the hydrazine was not occurred in the auxiliary water cooling system and vessel cooling system because of the coolant temperature was low. (3) An indistinct procedure is clarified and procedure of water quality analysis was established in the HTTR. (4) It is assumed that the corrosion of the components in these water cooling system hardly occurred from measurement results of dissolved oxide and chloride ion. Thus, the water quality was managed enough.

JAEA Reports

Improvement of helium sampling system in HTTR

Sekita, Kenji; Kuroha, Misao; Emori, Koichi; Kondo, Masaaki; Ouchi, Hiroshi; Shinozaki, Masayuki

JAEA-Technology 2008-002, 49 Pages, 2008/03

JAEA-Technology-2008-002.pdf:9.21MB

Graphite structures are used as one of the HTTR core internal structures. Graphite structures have high heat resistant property but its mechanical strength degrades easily by oxidization. To prevent the oxidization of graphite structures, impurity concentrations in the coolant of helium are controlled strictly. The helium sampling system is installed to measure the impurity concentrations in the helium. At gas compressor in helium sampling system, seal-oil leak at rod seal mechanism was occurred. The causes are degradation of seal material and contaminant abrasion powder of grand-packing. As these countermeasure, material of seal material was changed and contaminant was decreased. As the result long term operation is enabled. Moreover, reliable data can be obtained and efficient impurity control is enabled due to renewal of data acquisition control computer of gas chromatograph mass spectrometer and improvement of liquid nitrogen trap.

JAEA Reports

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.

JAEA Reports

Maintenance and management of emergency air purification system in HTTR

Aono, Tetsuya; Kondo, Masaaki; Sekita, Kenji; Emori, Koichi; Kuroha, Misao; Ouchi, Hiroshi

JAEA-Testing 2006-004, 39 Pages, 2006/06

JAEA-Testing-2006-004.pdf:9.88MB

The High Temperature Engineering Test Reactor (HTTR) has an emergency air purification system(EAPS). The system keeps the service area negative pressure condition and exhausts the filtered air to prevent fission products release to environment in accident condition. The EAPS is one of the engineered safety features which is started automatically when radioactivity in the service area increase or might increase. The performance of the EAPS should satisfy the analytical condition for public dose evaluation in the severest accidents of the HTTR. The performance should be confirmed by function tests. The function tests are divided into many tests corresponding to each assumed phenomenon. The confirmation of the performance of the system was carried out effectively by the tests. Moreover, the stable operation of the system can be achieved by improvements of the method of leak tight tests of exhaust filter unit. The report describes the outline of EAPS system, maintenance works and improvement of the system.

JAEA Reports

Maintenance of helium sampling system in HTTR

Sekita, Kenji; Emori, Koichi; Kuroha, Misao; Kimishima, Satoru; Wakabayashi, Hiroshi

JAEA-Testing 2006-001, 49 Pages, 2006/06

JAEA-Testing-2006-001.pdf:6.24MB

no abstracts in English

JAEA Reports

Report of investigation on malfunction of reserved shutdown system in HTTR

Hamamoto, Shimpei; Iigaki, Kazuhiko; Shimizu, Atsushi; Sawahata, Hiroaki; Kondo, Makoto; Oyama, Sunao; Kawano, Shuichi; Kobayashi, Shoichi; Kawamoto, Taiki; Suzuki, Hisashi; et al.

JAEA-Technology 2006-030, 58 Pages, 2006/03

JAEA-Technology-2006-030.pdf:10.69MB

During normal operation of High Temperature engineering Test Reactor (HTTR) in Japan Atomic Energy Agency (JAEA), the reactivity is controlled by the Control Rods (CRs) system which consists of 32 CRs (16 pairs) and 16 Control Rod Drive Mechanisms (CRDMs). The CR system is located in stand-pipes accompanied by the Reserved Shutdown System (RSS). In the unlikely event that the CRs fail to be inserted, the RSS is provided to insert B$$_{4}$$C/C pellets into the core. The RSS shall be designed so that the reactor should be held subcriticality from any operation condition by dropping in the pellets. The RSS consists of B$$_{4}$$C/C pellets, hoppers which contain the pellets, electric plug, driving mechanisms, guide tubes and so on. In accidents when the CRs cannot be inserted, an electric plug is pulled out by a motor and the absorber pellets fall into the core by gravity. A trouble, malfunction of one RSS out of sixteen, occurred during a series of the pre-start up checks of HTTR on February 21, 2005. We investigated the cause of the RSS trouble and took countermeasures to prevent the issue. As the result of investigation, the cause of the trouble was attributed to the following reason: In the motor inside, The Oil of grease of the multiplying gear flowed down from a gap of the oil seal which has been deformed and was mixed with abrasion powder of brake disk. Therefore the adhesive mixture prevented a motor from rotating.

JAEA Reports

Planning of the ATWS Simulation Tests by using Experimental Fast Reactor JOYO

Oyama, Kazuhiro; Kuroha, Takaya*; Takamatsu, Misao; Sekine, Takashi

JNC TN9410 2005-010, 57 Pages, 2005/03

JNC-TN9410-2005-010.pdf:1.51MB

A study of passive safety test using JOYO has been carried out to demonstrate the inherent safety of sodium cooled fast reactors. In this study, emphasis was placed on the improvement of the accuracy of plant kinetics calculations. The Mimir-N2 analysis code, developed to analyze JOYO plant kinetics, was selected as the standard code for predicting plant behavior during transients.Mimir-N2 was previously modified for MK-III core in 2001, 2002. Then, it implemented MK-III performance testing estimate analysis. The MK-III performance test included manual reactor shutdown test and loss of power supply test etc. as transient tests. In order to further improve the accuracy of the calculation, the Mimir-N2 heat transport system models of the reactor vessel upper plenum, the hot leg of secondary heat transport system and the dump heat exchanger were modified based on the results of the MK-III performance test in 2003.In this year, it stores up to get the prospect to have paid to the implementation of the UTOP, the ULOF test which is planned as the passive safety test, it evaluated about the plant structure and UTOP, the ULOF analysis for the parameter of which was the investing reactivity and so on by the Mimir-N2 analysis code. As a result, we could work out the testing condition which has prospect.

JAEA Reports

Modification of the Mimir-N2 Plant Dynamic Code based on JOYO MK-III Performance Test Results

Takamatsu, Misao; Kuroha, Takaya*; Yoshida, Akihiro

JNC TN9410 2004-005, 51 Pages, 2004/03

JNC-TN9410-2004-005.pdf:1.25MB

A study of passive safety test using JOYO has been carried out to demonstrate the inherent safety of sodium cooled fast reactors. In this study, emphasis was placed on the improvement of the accuracy of plant kinetics calculations. The Mimir-N2 analysis code, developed to analyze JOYO plant kinetics, was selected as the standard code for predicting plant behavior during transients. Mimir-N2 was Previously modified based on the data from plant characteristics and natural circulation tests during JOYO MK-II. Recently, the model of the core and the heat transport system of Mimir-N2 was upgraded to correspond to the modified heat transport system for MK-III. The MK-III performance test included manual reactor shutdown test and loss of power supply test etc. as transient tests. Although the sodium temperatures calculated by Mimir-N2 agreed well with the measurement results in the MK-III performance test, it was observed that the calculated sodium temperature descent rate at reactor vessel inlet and dump heat exchanger inlet etc. were slightly larger than measured. In order to further improve the accuracy of the calculation, the Mimir-N2 heat transport system models of the reactor vessel upper plenum, the hot leg of secondary heat transport system and the dump heat exchanger were modified based on the results of the MK-III performance test. As a result of the Mimir-N2 modification, the calculation results had improved agreement with the measurement results and it was confirmed that Mimir-N2 can accurately calculate plant behavior during transients such as reactor shutdown and loss of power supply.

JAEA Reports

An Investigation of fuel and fission product behavior in rise-to-power test of HTTR, 2; Results up to 30 MW operation

Ueta, Shohei; Emori, Koichi; Tobita, Tsutomu*; Takahashi, Masashi*; Kuroha, Misao; Ishii, Taro*; Sawa, Kazuhiro

JAERI-Research 2003-025, 59 Pages, 2003/11

JAERI-Research-2003-025.pdf:2.53MB

In the safety design requirements for the High Temperature Engineering Test Reactor (HTTR) fuel, it is determined that "the as-fabricated failure fraction shall be less than 0.2%" and "the additional failure fraction shall be small through the full service period". Therefore the failure fraction should be quantitatively evaluated during the HTTR operation. In order to measure the primary coolant activity, primary coolant radioactivity signals the in safety protection system, the fuel failure detection (FFD) system and the primary coolant sampling system are provided in the HTTR. The fuel and fission product behavior was evaluated based on measured data in the rise-to-power tests (1) to (4). The measured fractional releases are constant at 2$$times$$10$$^{-9}$$ up to 60% of the reactor power, and then increase to 7$$times$$10$$^{-9}$$ at full power operation. The prediction shows good agreement with the measured value. These results showed that the release mechanism varied from recoil to diffusion of the generated fission gas from the contaminated uranium in the fuel compact matrix.

JAEA Reports

Evaluation of core bowing reactivity of Joyo MK-III performance tests core

Takamatsu, Misao; Kuroha, Takaya*; Yoshida, Akihiro

JNC TN9400 2003-012, 38 Pages, 2003/03

JNC-TN9400-2003-012.pdf:2.03MB

A study of passive safety test using Joyo has been carried out to demonstrate the inherent safety of sodium cooled fast reactor. In this study, emphasis was placed on the improvement on feedback reactivity calculation accuracy. Especially, the core bowing reactivity has been one of the main targets of this study because it can be designed to provide negative reactivity in the event of power excursion. The core bowing reactivity was calculated by "MERBA" (MEchanical behavior and Reactivity shift caused by core Bowing Analysis code system for Joyo)" which had been developed for Joyo MK-II core. Through the operation of MK-II core, measured power reactivity coefficient shows a power dependency on reactor thermal power, which is considered to be caused by core bowing effect. In order to investigate the mechanism of this phenomenon, the measured power dependency on power reactivity coefficient were compared with the calculation by "MERBA". It was confirmed that the power dependency on the power reactivity coefficient of the MK-II core could be explained by a core bowing reactivity. In the upgraded MK-III core, neutron flux, the coolant temperature difference between reactor inlet and outlet, etc. will be changed and these may effect on the core bowing reactivity. Therefore, the core bowing reactivity of MK-III performance tests core was evaluated using "MERBA" which was modified to analyze MK-III core. As a result of "MERBA" calculation, the core bowing behavior of MK-III performance tests core was almost similar to MK-II core; the fuel subassembly was initially bent toward the core center at zero power, because the middle spacer pad of fuel subassembly was suppressed by reflectors which have rather large permanent distortion. During the thermal power increase, the fuel subassembly was bent to outward and the negative reactivity was inserted by the fuel movement to outward.

Oral presentation

Studies of passive safety tests by using the experimental fast reactor Joyo; Verification of Joyo plant dynamics analysis code Mimir-N2

Takamatsu, Misao; Kuroha, Takaya*; Aoyama, Takafumi

no journal, , 

Studies for passive safety tests to be conducted by using Joyo have been carried out to demonstrate the inherent safety of LMFBR. In these studies, emphasis has been placed on the improvement of the calculation accuracy of the feedback reactivity. Therefore, in the passive safety tests, the feedback reactivity measurement is planned under the condition simulating ATWS event. The Mimir-N2 which has been developed to analyze Joyo plant dynamic is used for predicting plant behavior during transients. In order to determine the passive safety tests plan, the improvement of the accuracy of Mimir-N2 was required. In this study, transient tests including manual reactor shutdown test, loss of power supply test and preliminary transient over power test were conducted to verify the updated model of the Joyo MK-III core and the heat transport system of Mimir-N2. As a result, a good agreement was obtained between calculated and measured sodium temperatures.

Oral presentation

Transient response test in Joyo

Kawahara, Hirotaka; Takamatsu, Misao; Aoyama, Takafumi; Kuroha, Takaya*

no journal, , 

no abstracts in English

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