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Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Nippon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00395_1 - 16-00395_9, 2017/04

no abstracts in English

Journal Articles

Safety evaluation of self actuated shutdown system for Gen-IV SFR

Saito, Hiroyuki*; Yamada, Yumi*; Oyama, Kazuhiro*; Matsunaga, Shoko*; Yamano, Hidemasa; Kubo, Shigenobu

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

A self-actuated shutdown system (SASS) is a passive device, which can detach a control rod for reactor shutdown in response to excessive increase in coolant temperature. Since a detachment temperature, which triggers release of a control rod, and a response time are identified as important parameters for validity analyses, this study focused on investigation of the response time and the detachment temperature, and safety analysis to see feasibility of the SASS in low power. For this purpose, design modifications were made to shorten the response time and three-dimensional thermal-hydraulic analysis in a low power operation was carried out in order to confirm the response time. The resulting detachment temperature level is lower than previous studies, leading to improved safety parameters. Based on improved parameters, a safety analysis to see feasibility of the SASS in low power was carried out. From this safety evaluation, it was confirmed that core damage can be prevented by the SASS with flow collector in the case of LOF type ATWS event.

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

Journal Articles

Prediction of fission product release during the LOFC experiments at the HTTR

Shi, D.*; Xhonneux, A.*; Ueta, Shohei; Verfondern, K.*; Allelein, H.-J.*

Proceedings of 7th International Topical Meeting on High Temperature Reactor Technology (HTR 2014) (USB Flash Drive), 11 Pages, 2014/10

Demonstration tests were conducted using the High Temperature Engineering Test Reactor (HTTR) in Oarai, Japan, to confirm the safety of HTGR technologies and assure the expected physical phenomena to occur under given conditions. As part of the OECD directed LOFC (loss of forced cooling) project, a series of three tests at the HTTR has been planned with tripping of all gas circulators while deactivating all reactor reactivity control to disallow reactor scram due to abnormal reduction of primary coolant flow rate. The tests fall into anticipated transient without scram (ATWS) with occurrence of reactor recriticality. The paper will describe the Source Term Analysis Code System (STACY) newly developed at the Research Center J$"u$lich and present the results of fission product behavior in the HTTR core under the LOFC test conditions. STACY encompasses the original verified and validated computer models for simulating fission product transport and release. For verification of the modernized and extended version, it was assured that results obtained with the original tools could be reproduced. One of the new features of STACY is its ability to also treat fuel compacts of (full) cylindrical or annular shape and a complete prismatic block reactor core, respectively, supposed sufficient input data be available. In the paper, calculations are based on time-dependent neutronics and fluid dynamics results obtained with the Serpent and MGT models.

Journal Articles

Safety demonstration tests using high temperature engineering test reactor

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Tachibana, Yukio; Sakaba, Nariaki; Iyoku, Tatsuo

Nuclear Engineering and Design, 233(1-3), p.301 - 308, 2004/10

 Times Cited Count:19 Percentile:19.84(Nuclear Science & Technology)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005, and the second phase tests will be carried out from FY 2006.

Journal Articles

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Tachibana, Yukio; Nakagawa, Shigeaki; Takeda, Takeshi; Saikusa, Akio; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Sawa, Kazuhiro; Iyoku, Tatsuo

Nuclear Engineering and Design, 224(2), p.179 - 197, 2003/09

 Times Cited Count:13 Percentile:30.43

no abstracts in English

Journal Articles

Irradiated fuel behavior under power oscillation conditions

Nakamura, Takehiko; Nakamura, Jinichi; Sasajima, Hideo; Uetsuka, Hiroshi

Journal of Nuclear Science and Technology, 40(5), p.325 - 333, 2003/05

 Times Cited Count:2 Percentile:78.8(Nuclear Science & Technology)

In order to examine high burnup fuel performance and to confirm its integrity under unstable power oscillation conditions arising during an ATWS in BWRs, two tests of irradiated fuels under simulated power oscillation conditions were conducted in the NSRR. Irradiated fuels at burnups of 25 and 56GWd/tU were subjected to four to seven power oscillations, which peaked at 50 to 95kW/m with intervals of 2s. The power oscillation was simulated by quick withdrawal and insertion of six regulating rods of the NSRR with a computerized control. Deformation of the fuel cladding of the test rods was comparable to those observed in shorter transient tests, which simulated RIAs, at the same fuel enthalpy level up to 368J/g. The fuel deformation was mainly caused by PCMI and was roughly proportional to the fuel enthalpy. Enhanced cladding deformation due to ratcheting by the cyclic load was not observed. Fission gas release, on the other hand, was considerably smaller than in the RIA tests, suggesting different release mechanisms in the two types of transients.

JAEA Reports

A Design study on a large FBR plant enhancing passive safety

Hayashi, Hideyuki; ;

PNC-TN9410 96-062, 186 Pages, 1996/02

PNC-TN9410-96-062.pdf:5.83MB

A conceptual design study on a 1300MWe large FBR plant was performed with focusing on enhancing passive safety and capital cost reduction. Spectrum-adjusted mixed nitride fueled core in which zirconium hydride was added, was applied to enlarge Doppler reactivity coefficient. Breeding ratio of 1.2 was obtained only with one layer of radial blanket subassemblies by optimizing the content of the zirconium hydride. The optimization also lightened the burden to the reactor structure through the reduction of the core diameter. Reactor passive shutdown were performed in the ATWS events of ULOF and ULOHS, and UTOP caused by one control rod full runout was endurable under the criterion of the prevention of coolant boiling. The safety feature can be called as inherent safety, because the feature comes only from the reactivity characteristics of the core. The integrities of the reactor structures which characterize head-access loop type reactor were evaluated on the transient thermal stress at the loss of flow accident and on seismic strain. Vertical strain of core support plate at loss of flow condition was also evaluated on the passive shutdown at ULOF. The capital cost of the large FBR plant was estimated 1.3 to 1.4 times as high as that of the same scale LWR based on the weight of major components.

JAEA Reports

None

PNC-TN1410 94-052, 181 Pages, 1994/06

PNC-TN1410-94-052.pdf:5.58MB

no abstracts in English

JAEA Reports

Critical heat flux for rod bundle under high-pressure boil-off conditions

Guo, Z.*; Kumamaru, Hiroshige; Kukita, Yutaka

JAERI-M 93-238, 20 Pages, 1993/12

JAERI-M-93-238.pdf:0.67MB

no abstracts in English

JAEA Reports

JAEA Reports

A Study on the reliability of the FBR reactor shutdown system; Design study for the large scale FBR

PNC-TN9410 91-286, 117 Pages, 1991/08

PNC-TN9410-91-286.pdf:9.35MB

A conventional type of RSS in a large scale FBR was designed and its unavailability was analyzed with fault-tree. Reliability of logic circuits of the reaetor protection system is relatively high when compared to that of the control rod insertion. Contributing factors to the unavailabity are multiple failures of detection systems, and failure to insert rods such as failure to deratch or rod jamming. Then the new concept of control rod release mechanism was introduced in the RSS design. The thermal-hydraulic characteristics of the mechanism was analyzed using computer codes SSC-L and AQUA. Further, qualitative analysis of the common cause failure for the RSS was tried with the generic cause approach. The reactor protection systems of the backup RSS are diversified by the self actuated control rod release mechanism. With such a mechanism, the number of common cause factors were decreased for postulated LOF event.

Journal Articles

Accident analyses for a double-flat-core type HCLWR

Okubo, Tsutomu; Iwamura, Takamichi; *; *; Murao, Yoshio

6th Proc. of Nuclear Thermal Hydraulics, p.79 - 86, 1990/11

no abstracts in English

Journal Articles

Accident analyses for a double-flat-core type HCLWR

Okubo, Tsutomu; Iwamura, Takamichi; *; *; Murao, Yoshio

Transactions of the American Nuclear Society, 62, p.662 - 663, 1990/11

no abstracts in English

JAEA Reports

Safety analysis of double-flat-core high conversion light water reactor; Large break LOCA and station blackout ATWS

*; Iwamura, Takamichi; Okubo, Tsutomu; ; Murao, Yoshio

JAERI-M 90-047, 37 Pages, 1990/03

JAERI-M-90-047.pdf:1.09MB

no abstracts in English

JAEA Reports

Key design parameter study (II) for large scale-up fast breeder reactor; Optimizing analysis of inherent negative reactivity feedback effect (I); Analysis on thermal transformation of core support plate

*; Tanigawa, Shingo*; *; *; *; *; *

PNC-TN9410 88-141, 159 Pages, 1988/09

PNC-TN9410-88-141.pdf:10.2MB

The structural analyses of the core support plate have been applied to study thermal transfomation behaviors and the differences of the movement by changing analytical model, under anticipated transient without scram (ATWS) conditions of FBR. The analyses have been performed for 1000 MWe class loop type fast breeder reactor using a structural analysis code FINAS. The thermal-hydraulic results, which have been performed to ATWS conditions using a plant system code, were used as the thermal boundary conditions to the calculation. The scope of the analyses included a whole section of reactor vessel and the dead load of core assemblies was also considered. Following results were obtained from these studies. (1)The thermal transformation of a upper core support plate can be evaluated according to the free expansion behavior owing to the temperature change of core support plate itself. (2)The radial restriction due to core subassemblies has much influence on the axial bend of the core support plate. (3)There are some differences to the transformation results between by the whole model and by the one dimensional model during the thermal transient is large. Another analysis will be needed, however, about the reactivity change according to the displacement of the core structure.

JAEA Reports

Key technological design study of a large LMFBR (I); Improvement of reactivity feedback modeling in SCC-L and analysis of plant thermal hydraulic behavior during ATWS accident

*; Ohshima, Hiroyuki

PNC-TN9410 88-006, 71 Pages, 1988/01

PNC-TN9410-88-006.pdf:9.72MB

Reactivity Feedback Modeling in Super System Code (SSC) has been improved to analyze the whole plant thermal hydraulic response to an anticipated transient without scram (ATWS) in a liquid metal fast breeder reactor (LMFBR). First of all, two-dimensional (2D) fluid flow and heat transfer modeling of reactor upper plenum (UP) has been modified. The heat transfer between the coolant and the control rod driveline (CRD) can be evaluated based on the 2D, not one-point, temperature distribution calculated by the UP model. The CRD is included as a part of in-plenum structure, and the thermal expansion of it is evaluated assuming the elongation is proportional to the temperature rise of the CRD. The reactivity feedback effect is evaluated using the elongation and the control rod worth. SSC-L is now capable of treating the following reactivity feedback effects caused by; (1)fuel doppler, (2)sodium density, (3)fuel axial expansion, (4)thermal expansion of the in-core structure, (5)thermal expansion of the core support structure, and (6)thermal expansion of the CRD. Whole plant thermal hydraulics during the ATWS accident can be analyzed taking the reactivity feedback effect into consideration more realistically than ever. An ATWS accident, i.e. unprotected loss-of-heat-sink (ULOHS), has been analyzed using SSC-L as an example. It is impossible to mitigate the ATWS consequence without core damage if no design change is made. However, it is found that more than 7 minutes of grace time is available for the remedial action at least if the above mentioned reactivity model is used. The accident progression is not so rapid in general in the ULOHS accident. The relatively slow response implies reactor shutdown can be achieved by a manual scram or a shutdown system actuated at slower speed can be utilized for mitigating the ATWS consequence.In the present analysis of CRD thermal expansion, it is assumed that the deformation of the CRD is one-dimensional, linear and elastic. It ...

JAEA Reports

Improvement of an advanced system code for loop-type lMFBRs, SSC-L; Modeling of reactivity feedback effects

Ohshima, Hiroyuki; *; Ninokata, Hisashi*

PNC-TN9410 87-122, 62 Pages, 1987/08

PNC-TN9410-87-122.pdf:3.15MB

In the safety analysis of ATWS (Anticipated Transient Without Scram) sequences, emphasis is placed on the thermohydraulics in reactor core and interactions between core and heat transport system (HTS) are not considered. However, if progress of the sequence is not so fast, thermohydraulics in reactor core and HTS should be calculated at the same time. A whole plant system code, such as SSC-L, is available for this purpose. Since SSC-L has not been applied to the ATWS analysis so far, reactivity feedback model in this code has something to be improved. If uncertainty in parameters used in the analysis is large, conservative assumptions are employed. Therefore, the results of simulations are conservative and have large uncertainty in general. In order to evaluate ATWS sequences, it is desirable to take reactor core and HTS interaction into consideration and to improve accuracy of reactivity feedback model in SSC-L as well as decreasing uncertainty of the input data. Grace period available for mitigating the ATWS by the operator recovery action can be also evaluated from the whole plant thermohydraulics. Therefore, reactivity calculation module in SSC-L has been modified and improved in this study. Thus the reactivity feedback effects calculated in SSC-L are as follows: (1)Fuel doppler, (2)Sodium density and void (3)Fuel axial expansion, (4)Thermal expansion of the core internal structure, and (5)Thermal expansion of the core support structure. For the purpose of checking performance of the new model, SSC-L has been applied to the simulations of ULOHS (Unprotected Loss of Heat Sink) accidents and the results are consistent in our perception. This model should be validated by experiments and SSC-L is to be extensively applied to the safety analysis of LMFBR plants in future.

Journal Articles

Improvement of passive safety of reactors

Asahi, Yoshiro; ; *

Nuclear Science and Engineering, 96, p.73 - 84, 1987/00

 Times Cited Count:3 Percentile:45.45(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Thermal-Hydraulic Behaviors during a loss of offsite power transient without SCRAM of a PWR

; ; ; ; ;

Nippon Genshiryoku Gakkai-Shi, 28(11), p.1045 - 1055, 1986/11

 Percentile:100(Nuclear Science & Technology)

no abstracts in English

30 (Records 1-20 displayed on this page)