• Nuclear Engineering and Technology
• /
• v.32 no.4
• /
• pp.395-409
• /
• 2000

Parametric studies were performed to assess the sensitivity in determining the maximum in-vessel heat removal capability from the core material relocated into the lower plenum of the reactor pressure vessel (RPV）during a core melt accident. A fraction of the sensible heat can be removed during the molten jet delivery from the core to the lower plenum, while the remaining sensible heat and the decay heat can be transported by rather complex mechanisms of the counter-current flow limitation (CCFL) and the critical heat flux (CHF）through the irregular, hemispherical gap that may be formed between the freezing oxidic debris and the overheated metallic RPV wall. It is shown that under the pressurized condition of 10MPa with the sensible heat loss being 50% for the reactors considered in this study, i.e. TMI-2, KORI-2 like, YGN-3&4 like and KNGR like reactors, the heat removal through the gap cooling mechanism was capable of ensuring the RPV integrity as much as 30% to 40% of the total core mass was relocated to the lower plenum. The sensitivity analysis indicated that the cooling rate of debris coupled with the sensible heat loss was a significant factor The newly proposed heat removal capability map (HRCM) clearly displays the critical factors in estimating the maximum heat removal from the debris in the lower plenum. This map can be used as a first-principle engineering tool to assess the RPV thermal integrity during a core melt accident. The predictive model also provided ith a reasonable explanation for the non-failure of the test vessel in the LAVA experiments performed at the Korea Atomic Energy Research Institute (KAERI), which apparently indicated a cooling effect of water ingression through the debris-to-vessel gap and the intra-debris pores and crevices.

• Environmental and Resource Economics Review
• /
• v.26 no.2
• /
• pp.139-172
• /
• 2017

Nuclear safety measures such as safety technology, culture, and regulation affects nuclear performances. This paper analyzes the change of nuclear performance by considering nuclear safety measures. Nuclear performance and technical data ranging 1970 to 2015 are collected from the Power Reactor Information System (PRIS) of IAEA. The result of panel regression analysis shows that overall engineering level, maintenance engineering and productivity decrease the forced loss rate (FLR). FLR structurally increase after Chernobyl accident in 1986 whereas after TMI and Fukushima accidents FLR didn't show any significant changes. The structural increase of FLR after Chernobyl are likely to result from the efforts of international communities for nuclear safety culture which makes nuclear operating company pay more opportunity cost to achieve nuclear safety.

• Nuclear Engineering and Technology
• /
• v.27 no.2
• /
• pp.203-215
• /
• 1995

A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called “Trip-Two/Leave-Two” (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3&4. The trip setpoint of the first too RCPs for YGN 3&4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft$^2$ break size in the hot leg. The analysis results show that YGN 3&4 can maintain the core coolability even if the operator fails to trip the second too RCPs or trips at worst time. Also, the YGN 3&4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3&4 can provide improved operator guidance for the RCP operation during accidents.

• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.625-636
• /
• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.