Kune Y. Suh

In-Vessel Retention of Molten Corium: Lessons Learned and Outstanding Issues

In-vessel retention (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 (LWRs). 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 accident. If it is possible to ensure that the vessel 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 advanced 600 MW(electric) pressurized water reactor (AP600) designed by Westinghouse, which relied upon external reactor vessel cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission approving the design without requiring that certain features common to existing LWRs, such as containment sprays, be safety related. Clearly, ERVC offers the potential to reduce the AP600’s construction and operating costs. However, it is not clear that the ERVC proposed for the AP600 could provide sufficient heat removal for higher-power reactors [up to 1500 MW(electric)] without additional enhancements. This paper reviews efforts made and results reported regarding the enhancement of IVR in LWRs. Where appropriate, the paper identifies what additional data or analyses are needed to demonstrate that there is sufficient margin for successful IVR in high-power thermal reactors.

THE CONCEPT OF PROLIFERATION-RESISTANT, ENVIRONMENT- FRIENDLY, ACCIDENT-TOLERANT, CONTINUAL AND ECONOMICAL REACTOR (PEACER)

In an effort to ameliorate generic concerns with current power reactors such as the risk of proliferation, radiological hazard of the spent fuel, and the vulnerability to core-melt accidents, the concept of a revolutionary reactor, named PEACER, has been developed as a proliferation-resistant waste transmutation reactor based on the unique combination of technologies of a proven fast reactor and the heavy liquid metal coolant. In this paper, results of the PEACER conceptual design are presented by focusing on the estimated performance of the PEACER system. The proliferation resistance of PEACER is based upon both institutional and technical issues. The latter includes denaturing of flssile materials, Pu in particular, as well as the intense radiation field associated with the pyrochemical partitioning method. When the fuel volume fraction and the core aspect ratio(L/D) are optimized, the transmutation capability of PEACER for long-lived wastes from LWR spent fuels is found to exceed the production rate of two LWR’s with the same electric rating. In contrast with current power reactor design principles, the lower power density and the higher neutron leakage rate lead to higher performance with respect to proliferation-resistance, transmutation capability and the accident-tolerance. Results of the present conceptual design show promising characteristics in all the five targets proposed by its name PEACER, which warrants more detailed study. 0 2000 Elsevier Science Ltd. All rights reserved.

Debris interactions in reactor vessel lower plena during a severe accident II. Integral analysis

Abstract The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 °C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 Vessel Inspection Program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation.

Debris interactions in reactor vessel lower plena during a severe accident. I. Predictive model

An integral predictive physico-numerical model has been developed to understand and interpret debris interactions in the reactor vessel plenum such as those which took place in the TMI-2 accident. The model represents the extent of debris jet disintegration by a jet-water entrainment model which can result in two types of debris configurations. One is particulated debris which eventually quenches in the water as a result of the entrainment process. The remainder of the debris penetrates to the bottom of the lower plenum and collects as a continuous layer. Each is treated as a separate region and has governing principles for its behavior. The potential for creating gap (contact) resistance and boiling heat removal is considered for heat transfer between the debris bed, the reactor vessel and steel structures and, most importantly, the vessel-to-crust gap water. The proposed in-vessel cooling mechanism due to material creep and water ingression into the expanding gap between the core debris and the vessel wall was found to explain the non-failure of the TMI-2 vessel in the course of the accident. The particulate debris bed is a mixture of metal and oxide, which is distributed as individual spherical particles of sizes determined at the time of entrainment. Energy is received from the continuum bed below by radiation and convection. The continuum debris bed is described by the crust behavior with the heat flux to the crust given by the natural convection correlations relating the Nusselt and Rayleigh numbers for the central region of debris. Using these governing principles, the rate laws for heat and mass transfer are formulated for each type of debris condition in the lower plenum. With the integration of the individual rates, the formation, growth and possible shrinkage of these regions are calculated. The potential reactor vessel breach is accounted for by considering the combined thermal and mechanical response of the vessel wall. The two-step failure model allows the vessel to fail at two different locations and at two different times.

Natural Convection Heat Transfer in Two-Dimensional Semicircular Slice Pool

This paper presents results from the Mini-SIGMA (Simulation of Internal Gravity-driven Melt Accumulation) tests concerned with high Rayleigh number turbulent natural convection in a molten pool. Tests are conducted to check on functionality of a custom-designed heater and to obtain correlations in terms of the Rayleigh and Nusselt numbers. The internal heating method is adopted in the tests utilizing the cable-type heater. This study concentrates on the thermal load, angular heat flux distribution, and temperature distribution inside the molten pool. The test section is a two-dimensional slice whose diameter, height, and thickness are 250 mm, 125 mm, and 50mm, respectively. The pools curved wall, with a 23 mm thick copper plate, is cooled by a regulated water loop. A water-cooling system is used to maintain the temperature of water surrounding the test section as constant as practicable with time. Four thin cable-type heaters, each with a diameter of 2.4 mm and a length of 900 mm, are used to simulate internal heating in the pool. They are uniformly distributed in the semicircular section to supply a maximum of 1 kW power to the pool. The Rayleigh number is obtained from these test data up to 1010.

Mixed convection friction factors and nusselt numbers in vertical annular and subchannel geometries

Analytical solutions are obtained for fully developed vertical laminar mixed convection flows within annular and conventional rod bundle subchannel geometries. Friction factors and Nusselt numbers are presented and fitted as functions of Grq/Re. A modified friction factor is defined to be used in applications where only bulk-averaged fluid temperatures are available, as in the case of lumped parameter analyses and most one-dimensional experiments. It is shown that the modified friction factor can vary significantly from the standard definition, which highlights the necessity of using the modified friction factor in analyses where the bulk density is used to calculate the gravity component of the axial pressure gradient. Finally, the present analysis is compared with experimental data available in the literature.

Numerical Simulation of Water-Based Alumina Nanofluid in Subchannel Geometry

Turbulent forced convection flow of Al2O3/water nanofluid in a single-bare subchannel of a typical pressurized water reactor is numerically analyzed. The single-phase model is adopted to simulate the nanofluid convection of 1% and 4% by volume concentration. The renormalization group k-e model is used to simulate turbulence in ANSYS FLUENT 12.1. Results show that the heat transfer increases with nanoparticle volume concentrations in the subchannel geometry. The highest heat transfer rates are detected, for each concentration, corresponding to the highest Reynolds number Re. The maximum heat transfer enhancement at the center of a subchannel formed by heated rods is ~15% for the particle volume concentration of 4% corresponding to Re = 80,000. The friction factor shows a reasonable agreement with the classical correlation used for such normal fluid as the Blasius formula. The result reveals that the Al2O3/water pressure drop along the subchannel increases by about 14% and 98% for volume concentrations of 1% and 4%, respectively, given Re compared to the base fluid. Coupled thermohydrodynamic and neutronic investigations are further needed to streamline the nanoparticles and to optimize their concentration.

Critical heat flux for downward facing boiling on a coated hemispherical surface

An experimental study was performed to investigate the effect of surface coating on the critical heat flux for downward facing boiling on the outer surface of a hemispherical vessel. Steady-state boiling experiments were conducted in the subscale boundary layer boiling (SBLB) facility using test vessels with metallic microporous coatings to obtain the local boiling curves and the local critical heat flux (CHF) limits. Similar heat transfer performance was observed for microporous aluminum and microporous copper coatings. When compared to the corresponding data without coatings, the boiling curves for the coated vessels were found to shift upward and to the right. This meant that the CHF limit was higher with surface coating and that the minimum film boiling temperatures were located at higher wall superheats. In particular, the microporous coatings were found to enhance the local CHF values appreciably at all angular locations explored in the experiments. Results of the present study showed that the microporous aluminum coating was very durable. Even after many cycles of steady state boiling, the vessel coating remained rather intact, with no apparent changes in color or structure. Although similar heat transfer performance was observed for microporous copper coatings, the latter were found to be much less durable and tended to degrade after several cycles of boiling.

Integral analysis of debris material and heat transport in reactor vessel lower plenum

Abstract An integral, fast-running, two-region model has been developed to characterize the debris material and heat transport in the reactor lower plenum under severe accident conditions. The debris bed is segregated into the oxidic pool and an overlying metallic layer. Debris crusts can develop on three surfaces: the top of the molten pool, the RPV wall, and the internal structures. To account for the decay heat generation, the crust temperature profile is assumed to be parabolic. The oxidic debris pool is homogeneously mixed and has the same material composition, and hence the same thermophysical properties, as the crusts, while the metallic constituents are assumed to rise to the top of the debris pool. Steady-state relationships are used to describe the heat transfer rates, with the assessment of solid or liquid state, and the liquid superheat in the pool being based on the average debris temperature. Natural convection heat transfer from the molten debris pool to the upper, lower and embedded crusts is calculated based on the pool Rayleigh number with the conduction heat transfer from the crusts being determined by the crust temperature profile. The downward heat flux is transferred to the lowest part of the RPV lower head through a crust-to-RPV contact resistance. The sideward heat flux is transferred to the upper regions of the RPV lower head as well as to the internal structures. The upward heat flux goes to the metal layer, water, or available heat sink structures above. Quenching due to water ingression is modeled separately from the energy transfer through the crust. The RPV wall temperature distribution and the primary system pressure are utilized to estimate challenges to the RPV integrity. Should the RPV be submerged, the heat removal is enhanced by the ex-vessel cooling due to nucleate boiling. The convection heat transfer correlations for the molten debris pool were validated against available experimental data and theoretical predictions. Testing of the model for a range of conditions in a PWR lower plenum produced consistent results. In addition, a comparison of the integral approach to a more detailed, special purpose model showed good agreement.

An Experimental Correlation of Cross-Flow Pressure Drop for Triangular Array Wire-Wrapped Rod Assemblies

An experimental study was carried out to quantitatively estimate the lateral drag changes due to flow structure alteration caused by the presence of wirewrap spacers in liquid-metal fast breeder reactor rod assemblies. Specially designed test rod bundles were constructed employing vertical straight wires attached at various angles around the rods relative to the cross-flow direction. These bundles simulate the cross-flow pressure drop within a control volume with axial mesh size less than one-twelfth of wire-wrap lead length. The variables examined were wire angular positions, Reynolds number, and rod arrangements. The transverse pressure drop data for triangular-array rod bundles with wires have been correlated throughout the laminar and turbulent flow regimes. The correlation is in the form of a correction parameter to be applied to the friction factor-Reynolds number relationship for the corresponding bare rod bundle.