16.2 - Severe Accident Analysis (RRC-B) - EDF Hinkley Point

SUB-CHAPTER : 16.2PRE-CONSTRUCTION SAFETY REPORTCHAPTER 16: RISK REDUCTION AND SEVEREACCIDENT ANALYSESPAGE : 77 / 295Document ID.No.UKEPR-0002-162 Issue 04Safety injection is assumed to be actuated at PZR pressure (< 112 bar); the safety injectionsignal automatically starts the MHSI+LHSI pumps and initiates a partial cooldown of thesecondary system. If the automatic partial cooldown is not successful, the operator will havetime (at least 30 minutes) to perform a manual partial cooldown. For this reason, it isassumed that preparation for primary Feed & Bleed starts 30 minutes after low PZR pressureand the identification of failure of partial cooldown.Two criteria can be used to decide to perform primary Feed & Bleed:• water level measurement (bottom of hot leg) and primary pressure• overheating at core outlet and secondary pressure2.2.3.1.3. Severe accident criteriaIf the PDS is not already actuated for primary Feed & Bleed, the ultimate criterion for theactuation of the PDS valves has the objective of depleting the accumulator tanks before theonset of core melt, in order to prevent potential high levels of oxidation due to accumulatorwater injection onto a hot core.This criterion (temperature of 650°C at the core outlet) is the ultimate criterion for openingone PDS line for the mitigation of oxidation. However, for all high pressure scenarios, theoperator is aware that at the time at which the criterion for primary Feed & Bleed is met hewill be asked to open one PDS line. Section 16.2.2.2 - Table 1 shows the evolution of theTLOFW scenario with late opening of PDS valves.2.2.3.2. Justification of PDS valve discharge capacity for in-vessel refloodingWith the strategy of opening PDS valves at the maximum core exit temperature of 650°C, thedischarge capacity of 900 te/hour at 176 bar for each PDS line (including two valves inseries) is sufficient to enable a very low pressure (below 4 bar) to be reached before theonset of relocation of molten core materials into the reactor vessel lower head, giving anadequate time margin before vessel failure. The differential pressure between theRCP [RCS] and the containment will thus be negligible at the time of reactor vessel failure(Section 16.2.2.2 - Figure 2).The discharge capacity of the PDS valves influences the pressure spike at the time of coriumrelocation into the lower head; this spike is in the range of 20 to 60 bar. This uncertaintydepends mainly upon corium fragmentation level and corium/water heat transfer whichdepends on assumptions used in the code. However, the high discharge capacity allows arapid decrease of pressure following this spike because the steam is easily removed. Thus,low pressure is reached early before vessel failure without any significant amount of water inthe lower head. The pressure difference with the containment is negligible at vessel failuretime. No risks of Direct Containment Heating (DCH) or significant uplift forces on the vesselsupports are expected. The corium release into the reactor pit will be a gravity driven flow.The main results of a core melt scenario LOOP with 6 EDG unavailable and opening of onePDS line at 650°C are given in Section 16.2.2.2 - Figures 2 to 8.