Nuclear Energy Research Initiative Research Articles

Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools.UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results.Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches.The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

Summary of the Sulfur-Iodine Process Integrated Laboratory-Scale Experiment

As part of the International Nuclear Energy Research Initiative project supported by the U.S. Department of Energy Office of Nuclear Energy, a collaborative team including Sandia National Laboratories, the Commissariat à l’Énergie Atomique in France, and industrial partner General Atomics constructed and operated a closed-loop system for demonstration of hydrogen production by the sulfur-iodine (S-I) process. The Integrated Laboratory-Scale experiment was conducted at General Atomics’ San Diego facility. This paper will summarize project goals, results of the program, key challenges identified for the S-I process, and the lessons learned.

Resonance Region Nuclear Data Analysis to Support Advanced Fuel Cycle Development

The Oak Ridge National Laboratory (ORNL) and the Korean Atomic Energy Research Institute (KAERI) are performing collaborative research as part of a three-year United States (U.S.)–Republic of Korea (ROK) International Nuclear Energy Research Initiative (I-NERI) project to provide improved neutron cross-section data with uncertainty or covariance data important for advanced fuel cycle and nuclear safeguards applications. ORNL and KAERI have initiated efforts to prepare new cross-section evaluations for Pu, Np, and the stable Cm isotopes. At the current stage of the I-NERI project, ORNL has recently completed a preliminary resonance-region cross-section evaluation with covariance data for Pu and has initiated resonance evaluation efforts for Np and Cm. Likewise, KAERI is performing corresponding high-energy cross-section analyses (i.e., above the resonance region) for the noted isotopes. The paper provides results pertaining to the new resonance region evaluation efforts with emphasis on the new Pu evaluation.

TRU-Fueled Very High Temperature Reactors for Applications Requiring an Extended Operation With Minimized Control and No Refueling

This paper presents an analysis of transuranium nuclide (TRU)-fueled very high temperature reactor (VHTR) systems focusing on applications requiring an extended operation with minimized control and no refueling (single-batch mode). As an example of such applications, international deployment opportunities for grid-appropriate VHTR systems could be mentioned addressing demands for electricity, industrial heat, and co-generation in those regions where minimized servicing is desirable for various reasons. The study is performed for the hexagonal block core concept within the framework of the ongoing U.S. DOE Nuclear Energy Research Initiative project on utilization of higher actinides (TRUs and partitioned minor actinides (MAs)) as a fuel component for extended-life VHTRs. The up-to-date analysis has shown reasonable reactivity swings, core life limits with respect to fast fluences, and criticality.

Autonomous Control Strategies for Very High Temperature Reactor Based Systems for Hydrogen Production

This paper is focused on feasible autonomous control strategies for Generation IV very high temperature reactors (VHTRs)-based systems for hydrogen production. Various burnable poison distributions and fuel compositions are considered. In particular, utilization of transuranium nuclides (TRUs) in VHTRs is explored as the core self-stabilization approach. Both direct cycle and indirect cycle energy conversion approaches are discussed. It is assumed that small-scale VHTRs may be considered for international deployment as grid-appropriate variable-scale self-contained systems addressing emerging demands for hydrogen. A Monte Carlo-deterministic analysis methodology has been implemented for coupled design studies of VHTRs with TRUs using the ORNL SCALE 5.1 code system. The developed modeling approach provides an exact-geometry 3D representation of the VHTR core details properly capturing VHTR physics. The discussed studies are being performed within the scope of the U.S. DOE Nuclear Energy Research Initiative project on utilization of higher actinides (TRUs and partitioned minor actinides) as a fuel component for extended-life VHTR configurations.

Development of GAMMA Code as an Integrated Code Analyzing a Coupled Very High Temperature Gas-Cooled Reactor and Hydrogen Production Plant

The very high temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 1173 K (900°C) and operational fuel temperatures above 1523 K (1250°C). The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the high-temperature gas-cooled reactor concepts have sufficiently high temperatures to support process heat applications, such as hydrogen production, tar sands, oil shale, desalination, or cogenerative processes, the VHTR’s higher temperatures can be detrimental to safety if a loss-of-coolant accident occurs and causes the mechanical strength degradation of the supporting graphite in the lower plenum. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion or density-gradient–driven stratified flow phenomena and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Therefore, there was a need to develop a computer code that can be used for VHTR air ingress-related graphite oxidation analyses. Prior to the start of the Republic of Korea/United States International Nuclear Energy Research Initiative collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate an air ingress phenomenon in the VHTR. Therefore, we have worked for the past 3 yr on developing and validating advanced computational methods for simulating air ingress in the VHTR. The Idaho National Laboratory is developing a system integration model of VHTR and hydrogen production plant. GAMMA code is being considered to be an integrated computer tool to analyze the thermal hydraulics of the coupled plant. Computer models for a high-temperature steam electrolysis (HTSE) process were developed and were implemented in an overall system process optimization code, HYSYS. The HTSE model will be implemented into GAMMA code as the integrated computer tool.This paper describes the governing equations and numerical methods used in GAMMA code and presents a portion of verification of the GAMMA code along with turbomachinery models and HTSE models that will be linked to GAMMA code.

Effect of Process Variables During the Head-End Treatment of Spent Oxide Fuel

The development of a head-end processing step for spent oxide fuel that applies to both aqueous and pyrometallurgical technologies is being performed by the Idaho National Laboratory, the Oak Ridge National Laboratory, and the Korean Atomic Energy Research Institute through a joint International Nuclear Energy Research Initiative. The processing step employs high temperatures and oxidative gases to promote the oxidation of UO2 to U3O8. Potential benefits of the head-end step include the removal or reduction of fission products as well as separation of the fuel from cladding. The effects of temperature, pressure, oxidative gas, and cladding have been studied with irradiated spent oxide fuel to determine the optimum conditions for process control. Experiments with temperatures ranging from 500 to 1250°C have been performed on spent fuel using either air or oxygen gas for the oxidative reaction. Various flow rates and applications have been tested with the oxidative gases to discern the effects on the process. Tests have also been performed under vacuum conditions, following the oxidation cycle, at high temperatures to improve the removal of fission products. The effects of cladding on fission product removal have also been investigated with released fuel under vacuum and high-temperature conditions. Results from these experiments will be presented as well as operating conditions based on particle size and decladding characteristics.

Spectrum shifting as a mechanism to improve performance of VHTRs with advanced actinide fuels

Reprocessing of spent LWR fuel is an intrinsic part of the closed fuel cycle. While current technologies treat recovered minor actinides as high level wastes, the primary objective of one of the U.S. DOE Nuclear Energy Research Initiative (NERI) projects is to assess the possibility, advantages and limitations of designing a single-batch (no-refueling) very high temperature reactor (VHTR) configuration that utilizes transuranic nuclides (TRU) as a fuel component. Since both VHTR core design concepts, pebble bed and prismatic block assembly, permit flexibility in component configuration, fuel utilization and management, it is possible to improve fissile properties by neutron spectrum shifting through configuration adjustments. The presented analysis is focused on the TRU-impact on the single-batch mode (no-refueling) VHTR core lifetime. As a result of the analysis, promising performance characteristics have been demonstrated. The TRU-core configurations are expected to be suitable for long-term autonomous operation without intermediate refueling.

Initial Results from Investigations to Enhance the Performance of High Temperature Irradiation-Resistant Thermocouples

New fuel, cladding, and structural materials offer the potential for safer and more economic energy from existing reactor and advanced nuclear reactor designs. However, insufficient data are available to characterize these materials in high temperature, radiation conditions. To evaluate candidate material performance, robust instrumentation is needed that can survive these conditions. However, traditional thermocouples either drift due to degradation at high temperatures (above 1100 °C) or due to transmutation of thermocouple components. Thermocouples are needed which can withstand both high temperature and high radiation environments. To address this instrumentation need, the Idaho National Laboratory (INL) recently developed the design and evaluated the performance of a high temperature radiation-resistant thermocouple that contains commercially-available alloys of molybdenum and niobium (Rempe, 2006). Candidate thermocouple component materials were first identified based on their ability to withstand high temperature and radiation. Then, components were selected based on data obtained from materials interaction tests, ductility investigations, and resolution evaluations. Results from long duration (over 4000 hours) tests at high temperatures (up to 1400 °C) and thermal cycling tests demonstrate the stability and reliability of the INL-developed design. Tests in INL’s Advanced Test Reactor (ATR) are underway to demonstrate the in-pile performance of these thermocouples. However, severalmore » options have been identified that could further enhance the lifetime and reliability of the INL-developed thermocouples, allowing their use in higher temperature applications (up to at least 1700 °C). A joint University of Idaho (UI) and INL University Nuclear Energy Research Initiative (UNERI) is underway to investigate these options and ultimately, provide recommendations for an enhanced thermocouple design. This paper presents preliminary results from this UI/INL effort. Results are reported from tests completed to evaluate the ductility, resolution, transient response, and stability of thermocouples made from non-commercially available alloys of molybdenum and niobium. In addition, this paper reports preliminary insights gained by comparing the performance of thermocouples fabricated with alternate techniques and geometries.« less

Introduction to the Annular Fuel Special Issue

This special issue of Nuclear Technology is devoted to evaluation of a new fuel design for the pressurized water reactor (PWR). It has a potential to increase the safety margin and power density of this most widely deployed type of power reactor. The idea is simple and involves increasing the surface-to-volume ratio of the fuel by adopting an annular geometry with internal and external cooling of the fuel. Undertaken with support from the initial phase of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy, the idea has undergone a thorough examination by researchers from Massachusetts Institute of Technology with industrial support from Westinghouse, Gamma Engineering, AREVA, and Atomic Energy of Canada Limited. This dedicated issue provides an inclusive technical review and economic assessment of the project in nine technical papers from the project team members. Hence, conclusions of this research have been documented in one archival location, which we hope will be of value to the technical community.The United States led the world in the development of the PWR, and this technology won over many countries, including some that began the nuclear energy era with focus on other concepts. Today the world appears ready for another phase of expansion in nuclear energy, which likely will be based in large part on PWR technology. While many aspects of the primary coolant system and even the containment design have been revamped in the recently developed Generation III reactors, the fuel has been left in its original form. Yet, it is the fuel performance that controls the power that can be extracted from the core and the response to transients, in other words, the economics and safety of the power plant. Therefore, the PWR fuel deserves a much more thorough examination for innovations that enable improved performance of nuclear energy in the next few decades.The annular fuel development for PWRs is an example of near-term research and development effort that has essentially been abandoned by both the U.S. government and industry. The U.S. government had decided that the PWR technology was mature and should not be given more development funding; the industry had decided that to change the fuel implies starting a new learning curve to ensure high degree of reliability, and until very recently had no intention of covering the cost of going after such fundamental changes in nuclear technology. Well, the market for nuclear has now expanded, and the investment in changes in the fuel that can bring about increased safety or economics can no longer be cast aside because of limited potential return on investment. Hopefully, fuel vendors and users will benefit from the extensive work documented in this issue.

Simulant melt experiments on performance of the in-vessel core catcher

In order to enhance the feasibility of in-vessel retention (IVR) of molten core material during a severe accident for high-power reactors, an in-vessel core catcher (IVCC) was designed and evaluated as part of a joint United States-Korean International Nuclear Energy Research Initiative (INERI). The proposed IVCC is expected to increase the thermal margin for success of IVR by providing an “engineered gap” for heat transfer from materials that relocate during a severe accident and potentially serving as a sacrificial material under a severe accident. In this study, LAVA-GAP experiments were performed to investigate the thermal and mechanical performance of the IVCC using the alumina melt as simulant. The LAVA-GAP experiments aim to examine the feasibility and sustainability of the IVCC under the various test conditions using 1/8th scale hemispherical test sections. As a feasibility test of the proposed IVCC in this INERI project, the effects of IVCC base steel materials, internal coating materials, and gap size between the IVCC and the vessel lower head were examined. The test results indicated that the internally coated IVCC has high thermal performance compared with the uncoated IVCC. In terms of integrity of the base steel, carbon steel is superior to stainless steel and the effect of bond coat is found to be trivial for the tests performed in this study. The thermal load is mitigated via boiling heat removal in the gap between the IVCC and the vessel lower head. The current test results imply that gaps less than 10 mm are not enough to guarantee effective cooling induced by water ingression and steam venting there through. Selection of endurable material and pertinent gap size is needed to implement the proposed IVCC concept into advanced reactor designs.

ICONE15-10842 INITIAL RESULTS FROM INVESTIGATIONS TO ENHANCE THE PERFORMANCE OF HIGH TEMPERATURE IRRADIATION-RESISTANT THERMOCOUPLES

Several options have been identified that could further enhance the lifetime and reliability of INL-developed thermocouples for in-pile testing, allowing their use in higher temperature applications (up to at least 1700℃ (3092^oF)). A joint University of Idaho (UI) and INL University Nuclear Energy Research Initiative (UNERI) is underway to investigate these options and, ultimately, provide recommendations for an enhanced thermocouple design. This paper presents preliminary results from this UI/INL effort. Results are reported from tests completed to evaluate the ductility, temperature resolution, transient response, and stability of thermocouples made from non-commercially available alloys of molybdenum and niobium. In addition, this paper reports preliminary insights gained by comparing the performance of thermocouples fabricated with various diameters.

Evaluation of Long-Life Transuranics Breakeven and Burner Cores for Waste Minimization in a PB-GCFR Fuel Cycle

Transuranics (TRU) breakeven and burner core designs have been studied for the Pebble-Bed Gas-Cooled Fast Reactor (PB-GCFR), which was developed under a 2-yr U.S. Department of Energy Nuclear Energy Research Initiative project. The issues of minimizing waste production, fuel cost, and burnup reactivity swing, and maximizing TRU burning have been investigated primarily from a neutronics viewpoint. For TRU breakeven cores, it was found that for the given core power [300 MW(thermal)] and power density (50 MW/m3), the lowest amount of radiotoxic TRU to be processed is obtained for a long-life (single-batch) core of 30-yr duration. Minimizing the TRU processed results in a minimization of the TRU losses that ultimately will have to be entombed in a geologic repository.The results show that the single-batch, long-life PB-GCFR could be designed to operate over a wide range of cycle lengths and fuel loadings. By modifying the TRU feed to have a higher minor actinide (MA) fraction than contained in light water reactor spent fuel, the burnup reactivity swing for the long-life core can be reduced significantly. With this approach, it is also possible to configure the long-life PB-GCFR core as a TRU burner using nonuranium fuel. A nonuranium fuel PB-GCFR with 24% plutonium and 76% MAs can operate for 17 full-power years and achieve 25% burnup with a reactivity swing of 3%Δk.

CHF enhancement by vessel coating for external reactor vessel cooling

In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding the reactor cavity during a severe accident. As part of a joint Korean–United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the subscale boundary layer boiling (SBLB) facility at the Pennsylvania State University using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady-state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady-state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.

Computational Analysis of the Thermal-Hydraulic Characteristics of the Encapsulated Nuclear Heat Source

The encapsulated nuclear heat source (ENHS) is a modular reactor that was selected by the 1999 U.S. Department of Energy Nuclear Energy Research Initiative program as a candidate Generation IV reactor concept. It is a fast neutron spectrum reactor cooled by lead-bismuth eutectic using natural circulation. One of the unique features of the ENHS is that the fission-generated heat is transferred from the primary coolant to the secondary coolant through rectangular intermediate heat exchanger (IHX) channels. The decay heat is removed by the reactor vessel auxiliary cooling system (RVACS).Events of protected loss of heat sink (PLOHS) and unprotected transient overpower (UTOP) have been analyzed for the ENHS using the CERES transient simulation code for liquid-metal-cooled reactors.It is found that the ENHS core is sufficiently cooled by the RVACS under the PLOHS condition. The core flow rate is affected by the growth and disappearance of temperature stratification in the primary plenum. It is also found that even under the inconceivable UTOP event considered, the ENHS reactor core is not catastrophically damaged. This is due to negative reactivity feedback from the radial expansion of the core, the grid plate, and the Doppler effect. The use of high-performance ferritic steel instead of HT-9 and proper design of the reactor control system could provide large safety margins against cladding damage.

Development of an Enhanced Core Catcher for Improving In-Vessel Retention Margins

In-vessel retention (IVR) of core melt that may relocate to the lower head of a reactor vessel is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for several advanced light water reactors. A U.S.-Korean International Nuclear Energy Research Initiative project has been initiated to explore design enhancements that could increase the margin for IVR for advanced reactors with higher power levels [up to 1500 MW(electric)]. As part of this effort, an enhanced in-vessel core catcher is being designed and evaluated. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary). The first is a base material that has the capability to support and contain the mass of core materials that may relocate during a severe accident; the second is an oxide coating on top of the base material, which resists interactions with high-temperature core materials; and the third is an optional coating on the bottom side of the base material to protect it from oxidation during the lifetime of the reactor. This paper summarizes results from the in-vessel core catcher design and evaluation efforts, focusing on recently obtained results from materials interaction tests and prototypic testing activities.

Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel

Framatome ANP, Sandia National Laboratories (SNL), Oak Ridge National Laboratory (ORNL), and the University of Florida are cooperating on the U.S. Department of Energy Nuclear Energy Research Initiative (NERI) project 2001-0124 to design, assemble, execute, analyze, and document a series of critical experiments to validate reactor physics and criticality safety codes for the analysis of commercial power reactor fuels consisting of UO2 with 235U enrichments ≥5 wt%. The experiments will be conducted at the SNL Pulsed Reactor Facility.Framatome ANP and SNL produced two series of conceptual experiment designs based on typical parameters, such as fuel-to-moderator ratios, that meet the programmatic requirements of this project within the given restraints on available materials and facilities. ORNL used the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) to assess, from a detailed physics-based perspective, the similarity of the experiment designs to the commercial systems they are intended to validate. Based on the results of the TSUNAMI analysis, one series of experiments was found to be preferable to the other and will provide significant new data for the validation of reactor physics and criticality safety codes.

Conceptual design of an in-vessel core catcher

An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (US)–Korean International Nuclear Energy Research Initiative (INERI) investigating methods to insure retention of materials that may relocate to the lower head of a reactor vessel under severe accident conditions in advanced reactors. This enhanced core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulator coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. This paper summarizes results from thermal, flow, and structural analyses as well as initial scoping materials interaction tests that were completed to support the conceptual design of the core catcher.

Corium retention for high power reactors by an in-vessel core catcher in combination with External Reactor Vessel Cooling

If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, IVR of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.

Late-phase melt conditions affecting the potential for in-vessel retention in high power reactors

If cooling is inadequate during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 accident. In such a case, concerns about containment failure and associated risks can be eliminated if it is possible to ensure that the lower head remains intact so that relocated core materials are retained within the vessel. Accordingly, in-vessel retention (IVR) of core melt as a key severe accident management strategy has been adopted by some operating nuclear power plants and planned for some advanced light water reactors. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements can provide sufficient heat removal to assure IVR for high power reactors (i.e., reactors with power levels up to 1500 MWe). Consequently, a joint United States/Korean International Nuclear Energy Research Initiative (I-NERI) has been launched to develop recommendations to improve the margin of success for in-vessel retention in high power reactors. This program is initially focussed on the Korean Advanced Power Reactor—1400 MWe (APR1400) design. However, recommendations will be developed that can be applied to a wide range of existing and advanced reactor designs. The recommendations will focus on modifications to enhance ERVC and modifications to enhance in-vessel debris coolability. In this paper, late-phase melt conditions affecting the potential for IVR of core melt in the APR1400 were established as a basis for developing the I-NERI recommendations. The selection of ‘bounding’ reactor accidents, simulation of those accidents using the SCDAP/RELAP5-3D © code, and resulting late-phase melt conditions are presented. Results from this effort indicate that bounding late-phase melt conditions could include large melt masses (>120,000 kg) relocating at high temperatures (∼3400 K). Estimated lower head heat fluxes associated with this melt could exceed the maximum critical heat flux, indicating additional measures such as the use of a core catcher and/or modifications to enhance external reactor vessel cooling may be necessary to ensure in-vessel retention of core melt.