Investigation of thermal expansion model for evaluation of core support plate reactivity in ATWS event

Sotsu, Masutake

Journal of Energy and Power Engineering, 14(8), p.251 - 258, 2020/08

https://doi.org/10.17265/1934-8975/2020.08.001

Thermal expansion behavior was investigated for evaluation of the core support plate expansion reactivity in the Unprotected Loss of Heat Sink reactor trip failure event. A possibility of mechanical restraint was investigated in thermal expansion of the core structure for the prototype fast breeder reactor Monju. The reactor core expansion was simulated in a three-dimensional finite element analysis model of the reactor vessel considering detailed temperature distribution of the sodium coolant based on the thermal-hydraulic analysis result of the whole core model. It was found that the thermal expansion of the core was not restrained in the ULOHS evert, although part of the core structure is mechanically restrained.

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.

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 Jlich 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.

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

https://doi.org/10.1016/S0029-5493(03)00108-0

no abstracts in English

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.

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

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.

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

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*; *; Yamaguchi, Katsuhisa; *; *; *

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.

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.