Comparison on safety features among HTGR's Reactor Cavity Cooling Systems (RCCSs)

Takamatsu, Kuniyoshi; Funatani, Shumpei*

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 17 Pages, 2023/04

The objectives of this study are as follows: to understand the characteristics, degree of passive safety features for heat removal were compared for RCCSs based on atmospheric radiation and based on atmospheric natural circulation under the same conditions. Therefore, the authors concluded that the proposed RCCS based on atmospheric radiation has the advantage that the temperature of the RPV can be stably maintained against disturbances in the outside air (ambient air). Moreover, methodology to utilize all the heat emitted from the RPV surface for increasing the degree of waste-heat utilization was discussed.

Estimation of the core degradation and relocation at the Fukushima Daiichi Nuclear Power Station Unit 2 based on RELAP/SCDAPSIM analysis

Madokoro, Hiroshi; Sato, Ikken

Nuclear Engineering and Design, 376, p.111123_1 - 111123_15, 2021/05

https://doi.org/10.1016/j.nucengdes.2021.111123

Comparative methodology between actual RCCS and downscaled heat-removal test facility

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 133, p.830 - 836, 2019/11

https://doi.org/10.1016/j.anucene.2019.07.025

A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. Moreover, the authors started experiment research with using a scaled-down heat-removal test facility. Therefore, this study propose a comparative methodology between an actual RCCS and a scaled-down heat-removal test facility.

Experimental study on heat removal performance of a new Reactor Cavity Cooling System (RCCS)

Hosomi, Seisuke*; Akashi, Tomoyasu*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Takamatsu, Kuniyoshi

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11

A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We started experiment research with using a scaled-down test section. Three experimental cases under different emissivity conditions were performed. We used Monte Carlo method to evaluate the contribution of radiation to the total heat released from the heater. As a result, after the heater wall was painted black, the contribution of radiation to the total heat could be increased to about 60%. A high emissivity of RPV surface is very effective to remove more heat from the reactor. A high emissivity of the cooling part wall is also effective because it not only increases the radiation emitted to the ambient air, but also may increase the temperature difference among the walls and enhance the convection heat transfer in the RCCS.

An Evaluation study of measures for prevention of Re-criticality in sodium-cooled large FBR with MOX fuel

JNC TN9400 2000-038, 98 Pages, 2000/04

JNC-TN9400-2000-038.pdf:7.49MB

As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.

None

Mishima, Kaichiro*; Hibiki, Takashi*; *; Tobita, Yoshiharu

PNC TY9604 96-003, 10 Pages, 1996/05

PNC-TY9604-96-003.pdf:0.34MB

no abstracts in English

Comparative study of source terms of a BWR severe accident by THALES-2, STCP and MELCOR

Hidaka, Akihide; *; Soda, Kunihisa; Muramatsu, Ken; Sakamoto, Toru*

ANS Proc. of the 1992 National Heat Transfer Conf., p.408 - 416, 1993/00

no abstracts in English

None

PNC TN1410 92-026, 113 Pages, 1992/01

PNC-TN1410-92-026.pdf:11.01MB

no abstracts in English

The plant thermohydraulic analysis for the monju PRA study; Recovery from PLOHS or LORL using the maintenance cooling system

*; *

PNC TN9410 88-055, 111 Pages, 1988/06

PNC-TN9410-88-055.pdf:5.87MB

In this study, decay heat removal capability of the Maintenance Cooling System (MCS) of Monju has been investigated with respect to protected accidents. The protected accidents of the Liquid Metal Fast Breeder Reactors (LMFBRs), such as Protected Loss-of-Heat-Sink (PLOHS) or Loss-of-Reactor-Level (LORL), are of great importance from the viewpoint of the annual frequency of core damage. The progression of the protected accidents is mild in general because reactor decay heat can be dispersed from the core by natural circulation. The decay heat for Monju is to be removed by the Intermediate Reactor Auxiliary Cooling system (IRACS). It is essential to keep the intactness of coolant flow path from the reactor core to the heat sink and the availability of heat sink itself. If the either of them is degraded, it is taken for granted d that protected slow meltdown follows. However, the reactor core can be prevented from any damage or meltdown if the decay heat can be removed through MCS. The plant thermohydraulics of the procected accidents is analyzed using SSC-L to develop success criteria in the decay heat removal by the MCS. Parametric calculations are performed with respect to: available heat capacity in the heat transport system, cooling time before the loss-of-heat-sink and MCS starting time. It has been found, for example, that (1)MCS can remove the decay heat immediately after the reactor shutdown if heat capacity of more than two main coolant loops is available; (2)after two hours cooling time by natual circulation, MCS can remove the decay heat even if no coolant flow is assumed in all the main heat transport system; (3)LORL caused by the failure in sodium make-up can be recovered by the MCS operation. In the PLOHS condition, the coolant temperature may exceed conservative design limit of the MCS piping. However, the conservativeness of the design limit and the method of qualification make compensation for the deterioration in structural strength. Finally, ...

Sensitivity study on PWR source terms with THALES/ART code package and effects of in-vessel thermal-hydraulic models

Abe, Kiyoharu; Watanabe, Norio; Ida, Mitsuo; Sakamoto, Toru; Ishigami, Tsutomu; Harami, Taikan

Proc.on Probabilistic Safety Assessment and Risk Management PSA87, Vol.3, p.945 - 950, 1987/00

no abstracts in English