Canada has participated in the Generation IV International Forum (GIF) collaboration in the area of Super Critical Water-cooled Reactors (SCWR). Similar to the current CANDU technologies, in the Canadian SCWR design the low pressure heavy water moderator system is separated from the supercritical coolant system (25MPa). The High Efficiency Re-entrant Channel (HERC) design in the Canadian SCWR has multiple coolant regions (i.e. coolant in the downward center flow tube and coolant in the upward outer fuel region) and provides thermal isolation between the moderator and heat transport system fluid. Although the overall reactivity feedback in the channel is negative for equilibrium density decrease transients, a temporary positive reactivity may be induced during non-equilibrium conditions such as cold-leg Loss of Coolant Accidents (LOCA). The primary objective of this study is to investigate the fuel and coolant behaviors under postulated LOCA transients, in particular those caused by cold-leg breaks, and to demonstrate the effectiveness of several proposed safety systems in the Canadian SCWR.
The one-dimensional thermalhydraulic system code RELAP5 has been used for the safety analysis in many LWRs. The latest version RELAP5/MOD4 has been improved to accommodate supercritical water and is used in this study. A RELAP5 model is constructed based on the most recent Canadian SCWR design. The 336 fuel channels are split into two representative groups each with a series of hydraulic and heat structure models. A benchmark study is conducted by comparing RELAP5 to CATHENA simulations and shows good agreement for both steady-state and transient predictions.
The RELAP5 model is then used to predict the system response to several postulated LOCA transients. For a 100% (single-ended) cold-leg break located in between the feedwater pump and the inlet plenum, the system pressure immediately drops followed closely by a flow reversal with rapid discharge from the break. A brief power pulse (178%FP) is observed under this non-equilibrium depressurization scenario. The transient simulations show the potential for two sheath temperature maxima, one early in the transient as a result of the power pulse and the subsequent flow-power mismatch, and another later peak resulting from the fuel heat-up under near stagnant channel flow conditions (such as in the failure of the Emergency Core Cooling Systems) as the heat transfer regime changes to radiation dominated. The Automatic Depressurization System (ADS) located on the hot-leg side mitigates the later fuel heat-up by introducing forward channel flows. This effect is enhanced by additional coolant supplied from Low Pressure Coolant Injection (LPCI) which is part of the Emergency Core Cooling System (ECCS). Under the 100% break LOCA/LOECC transient, the core inventory is depleted rapidly after the break and thermal radiation becomes the dominant heat removal mechanism. The highest MCST, 1331 K, is achieved approximately 136s after the break and meets the safety criterion (1533 K). Beyond this time the sheath temperatures gradually decrease either by the continuous LPCI from the reactor sumps, gravity driven core cooling, or in the event of a failure of those systems by the Passive Moderator Cooling System (PMCS).
LOCAs initiated by break sizes varying from 5% to 100% of the cold-leg cross-section area are simulated under loss of ECCS. In this specific design, break sizes less than 15% are defined as SBLOCAs and show an early pressure increase up to the Safety Relief Valve (SRV) setpoint. During SBLOCAs, the first MCST peak is more limiting than the large LOCA case because of insufficient fuel cooling caused by relatively low reverse flow. However, these lower reverse flows prolong the period of blowdown cooling and hence help to mitigate the secondary MCST peak. The worst LOCA case occurs in the 15% break case with a maximum cladding temperature of 1450 K. The results showed the most sensitive parameters are delays associated with SDS action, emissivity and ADS actuation parameters. / Thesis / Master of Applied Science (MASc)
| Identifer | oai:union.ndltd.org:mcmaster.ca/oai:macsphere.mcmaster.ca:11375/19137 |
| Date | January 2016 |
| Creators | Lou, Mengmeng |
| Contributors | Novog, David, Engineering Physics and Nuclear Engineering |
| Source Sets | McMaster University |
| Language | English |
| Detected Language | English |
| Type | Thesis |
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