Report from JRR-3

Shibayama, Mitsuhiro*; Murayama, Yoji; Takeda, Masayasu

AONSA Newsletter (Internet), 11(1), P. 20, 2019/08

no abstracts in English

Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA Project

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

Journal of Power and Energy Systems (Internet), 3(1), p.146 - 157, 2009/00

https://doi.org/10.1299/jpes.3.146

Presented are experiment results of the LSTF with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break LOCA simulation experiment. The break size is equivalent to 1.9% cold leg break. The accident management (AM) action to rapidly open the SG relief valves was initiated when CET temperature rose up to 623 K. The core overheat, however, was detected with a time delay of about 230 s and a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarified the reasons of time delay and temperature discrepancy between the CETs and heated core including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to PWR conditions and a possibility of alternative indicators for earlier AM action is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

A Study on timing of rapid depressurization action during PWR vessel bottom break LOCA with HPI failure and AIS-gas inflow, ROSA-V/LSTF test SB-PV-06

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAEA-Research 2007-037, 150 Pages, 2007/03

JAEA-Research-2007-037.pdf:7.55MB

A small break LOCA experiment (SB-PV-06) was conducted at the LSTF of ROSA-V program to study effects of rapid secondary depressuriza-tion action on core cooling as one of accident management (AM) measures for a PWR in case of high pressure injection system failure and non-condensable gas inflow from the accumulator injection system. The break simulated 10 instrument tubes rupture equivalent to 0.2% cold leg break. It was clarified through comparison with former experiments that (1) the depressurization initiated by detecting the vessel level below the primary loop (4545s) was degraded by the gas inflow resulting in whole core uncovery prior to the start of low pressure injection and (2) an alternative start of the depressurization by detecting level decrease at the SG outlet plenum (2330s), would limit the core uncovery suggesting more effective parameter for the AM measures. The report presents the experiment results with the effects of rapid depressurization initiation timing.

An Experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow, ROSA-V test SB-PV-04

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAEA-Research 2006-018, 140 Pages, 2006/03

JAEA-Research-2006-018.pdf:7.14MB

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection system during an SBLOCA at a pressurized water reactor (PWR). The experiment (SB-PV-04) simulating a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes clarified that rapid depressurization action by full-opening of relief valves and supplying auxiliary feedwater were effective to avoid core uncovery through actuation of low pressure injection system irrespective of significantly degraded depressurization by non-condensable gas inflow from the accumulator tanks. It is clarified that the effective core cooling was established by the rapid primary cooling which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as another AM action and resulted in core heatup.

Effects of secondary depressurization on core cooling in PWR vessel bottom small break LOCA experiments with HPI failure and gas inflow

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

Journal of Nuclear Science and Technology, 43(1), p.55 - 64, 2006/01

https://doi.org/10.3327/jnst.43.55

Effects of non-condensable gas from the accumulator tanks on secondary depressurization, as one of accident management (AM) measures in case of high pressure injection system failure, are studied at the ROSA-V/LSTF experiments simulating a ten instrument-tube break LOCA at the PWR vessel bottom. In an experiment with no gas inflow, the secondary depressurization at -55 K/h initiated by SI signal with 10 minutes delay succeeded in the LPI actuation. On the other hand, the gas inflow in another experiment degraded the primary depressurization and resulted in core uncovery before the LPI start. The third experiment with rapid secondary depressurization and continuous auxiliary feedwater supply, however, showed a possibility of long-term core cooling by the LPI actuation. RELAP5/MOD3 code analyses well predicted these experiment results and clarified that condensation heat transfer was largely degraded by the gas in the U-tubes. In addition, a primary pressure - coolant mass map was found to be useful for indication of key plant parameters of AM measures.

Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

JAERI-Research 2005-014, 170 Pages, 2005/06

JAERI-Research-2005-014.pdf:7.64MB

A small break LOCA (SBLOCA) experiment was conducted at the LSTF of ROSA-V program to study effects of accident management (AM) on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a PWR. The experiment, SB-PV-03, simulated ten instrument-tube break LOCA at the PWR vessel bottom equivalent to 0.2% cold leg break, total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and AM actions on secondary depressurization at -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes. It was clarified that the AM actions were effective on primary depressurization until AIS injection end at 1.6 MPa, but thereafter became less effective by the gas inflow, resulting in low pressure injection (LPI) delay and whole core heatup under continuous water discharge at the break. The report describes these phenomena including core heatup related with primary coolant mass and AM actions, primary-to-secondary heat transfer analysis and estimation of gas in the primary loops.

Effects of secondary depressurization on PWR bottom small break LOCA experiments in case of HPI failure and non-condensable gas inflow

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10

no abstracts in English

Estimates of parameter and scenario uncertainties in shallow-land disposal of uranium wastes using deterministic and probabilistic safety assessment models

Takeda, Seiji; Kanno, Mitsuhiro; Minase, Naofumi; Kimura, Hideo

Journal of Nuclear Science and Technology, 39(8), p.929 - 937, 2002/08

https://doi.org/10.1080/18811248.2002.9715278

no abstracts in English

Effect of stress and impurities on preferential amorphization on grain boundaries in polycrystalline silicon

Takeda, Mitsuhiro*; Onuki, Somei*; Watanabe, Seiichi*; Abe, Hiroaki; Naramoto, Hiroshi; P.R.Okamoto*; N.Q.Lam*

Mat. Res. Soc. Symp. Proc., 540, p.37 - 42, 1999/00

no abstracts in English

Effects of three-dimensional steam flow on exit temperatures during core boil-off in PWR vessel top break LOCA simulation experiments by using ROSA/LSTF

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

no journal, , 

A core fluid temperature distribution which is considered to be affected by both high-temperature steam flow towards the control rod cluster guide tubes and low-temperature steam flow affected by non-heating structural materials especially in the core peripheral regions was detected during a core boil-off condition in PWR vessel top break LOCA simulation experiments conducted at ROSA/LSTF to study effectiveness of accident management measures in the OECD/NEA ROSA project. Reported are these three-dimensional steam flows in the core and their influences on detection of the super-heated temperatures at the core exit.