Potential for AP600 in-vessel retention through ex-vessel flooding

Abstract

This report was prepared as an account of work sponsored by an agency of the United States Government Neither the United States Government nor any agency thereof, nor any of their cmpioycu, makes any warranty, express or implied. or ujumts any legal liability or rrsponsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its w would not infringe privately owned rights. Reference herein to any specific commercial product, process, or Icrvicc by trade name, trademark, manufac-Cum, or othemke doej not necessarily constitute or imply its endorsement, ncommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors c x p d hemn do not nrmcarily state or reflect thosc of the United States Government or.any agency thereof.. . Contents Department of Energy (DOE) contracted the University of California-Santa Barbara (UCSB) to produce the peer-reviewed report, DOE/ID-10460.2 To assist in the Nuclear Regulatory Commission's (NRC's) evaluation of IVR of core melt by exvessel flooding of the AP600, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform the following: ERVC effectiveness. method to identify areas where technical concerns weren't addressed. margins to failure and conclusions presented in the UCSB study. An in-depth critical review of the UCSB study and the model that UCSB used to assess An in-depth review of the UCSB study peer review comments and of UCSB's resolution An independent analysis effort to investigate the impact of residual concerns on the This Technical Evaluation Report (TER) summarizes results from these tasks. As discussed in Sections 1.1 and 1.2, INEEL'S review of the UCSB study and peer reviewer comments suggested that additional analysis was needed to assess: (1) the integral impact of peer reviewer-suggested changes to input assumptions and uncertainties and (2) the challenge present by other credible debris configurations. NRC tasked INEEL to perform an independent assessment to address this need. Section 1.3 summarizes the corresponding analysis approach developed by INEEL. The remainder of this report provides more detailed descriptions of INEEL's analysis methodology, input assumptions, and results. 1.1 Review of UCSB Study and Model As noted in Reference 2, the objective of the UCSB study was to demonstrate the effectiveness of ERVC for an AP600-like design and to provide a readily adaptable path for demonstrating ERVC for other reactor designs. Figure 1-1 illustrates the UCSB approach for demonstrating AP600 vessel integrity for cases with complete RCS depressurization and ERVC. As indicated in this figure, the UCSB study attempts to demonstrate vessel integrity by proving two assertions: Assertion 1: For all heat fluxes at or below the critical heat flux (0, the corresponding Assertion 2: Heat fluxes from relocated melt to the lower head always remain below the CHF. minimum vessel wall thicknesses are sufficient that the...

Development of Thermodynamic Database for U-Zr-Fe-O-B-C-FPs System

“…Therefore, when some uranium in the oxide corium transfers into the metal layer, density inversion between the oxide and the metal could occur. The possibility of the density inversion was also indicated previously 5) based on experimental data 6) that suggested some uranium could dissolve into unoxidized zirconium. Therefore, the material effect on molten corium such as the carbon influence and the uranium transfer has become important for IVR estimation.…”

Investigations on in-vessel melt retention by external cooling for a generic VVER-1000 reactor

“…A value of 0.45 has been applied by (Theofanous et al, 1996). In (Rempe et al, 1997) a value of 0.29 is used. The influence of the emissivity is discussed separately in Section 5.2.…”

“…The in-vessel melt retention by ex-vessel cooling concept is described in (Theofanous et al, 1996(Theofanous et al, , 1997aRempe et al, 1997Rempe et al, , 2008. This strategy has been approved by the Finnish Regulatory Agency (STUK) to be a part of the severe accident management procedures for the Loviisa NPP with a VVER-440 reactor .…”

“…The Idaho National Laboratory (INL), USA, has developed the computer code VESTA. It was applied in a verification study for assessment of the potential of AP600 in-vessel retention through ex-vessel flooding (Rempe et al, 1997). At the Royal Institute of Technology (KTH), Sweden, the MVITA code was developed (Bui et al, 1998;Sehgal et al, 1999Sehgal et al, , 2000, which applies the effective conductivity-convectivity model (ECCM) of natural convection heat transfer in internally heated pools.…”

A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

“…A common practice in the IVR studies [1,6,7] is to determine the corium pool thermal load in the lower head using stationary configurations, initial corium inventory and arbitrary assumptions. This corium inventory is evaluated using integral codes which simulate core degradation such as MAAP4 [8] or MELCOR [9].…”