J. L. Rempe

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.

Thermocouples for High-Temperature In-Pile Testing

Traditional methods for measuring in-pile temperatures degrade above 1100°C. Hence, the Idaho National Laboratory (INL) initiated a project to explore the use of specialized thermocouples for high temperature in-pile applications. Efforts to develop, fabricate, and evaluate specialized high-temperature thermocouples for in-pile applications suggest that several material combinations are viable. Tests show that several low-neutron cross-section candidate materials resist material interactions and remain ductile at high temperatures. In addition, results indicate that the candidate thermoelements have a thermoelectric response that is single-valued and repeatable with acceptable resolution. The selection of the thermocouple materials depends on desired peak temperature and accuracy requirements. For applications at or above 1600°C, tests indicate that thermocouples having doped molybdenum and Nb-1%Zr thermoelement wires, HfO2 insulation, and a Nb-1%Zr sheath could be used. INL has worked to optimize this thermocouple’s stability. With appropriate heat treatment and fabrication approaches, results indicate that thermal cycling effects on this thermocouple’s calibration is minimized. INL initiated a series of high-temperature (1200 to 1800°C) long-duration (up to 6 months) tests to assess the long-term stability of these thermocouples. Initial results indicate that the INL-developed thermocouple’s thermoelectric response is very stable. Typically, <20°C drift was observed in a 4000-h test at 1200°C. In comparison, commercially available types K and N thermocouples included in these 1200°C tests experienced drifts up to 110°C.

Options Extending the Applicability of High-Temperature Irradiation-Resistant Thermocouples

Abstract Several options have been identified that could further enhance the reliability and extend the applicability of high-temperature irradiation-resistant thermocouples (HTIR-TCs) developed by the Idaho National Laboratory (INL) for in-pile testing, allowing their use in temperature applications as high as 1800%C.The INL and the University of Idaho (UI) investigated these options with the ultimate objective of providing recommendations for alternate thermocouple designs that are optimized for various applications. This paper reports results from INL/UI investigations. Results are reported from tests completed to evaluate the ductility, resolution, transient response, and stability of thermocouples made from specially formulated alloys of molybdenum and niobium,not considered in initial HTIR-TC development. In addition, this paper reports insights gained by comparing the performance of HTIR-TCs fabricated with various heat Ntreatments and alternate geometries.

Review Paper: Review of Instrumentation for Irradiation Testing of Nuclear Fuels and Materials

Abstract Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in material test reactors (MTRs). Recently, there is increased interest to irradiate new materials and reactor fuels for advanced pressurized water reactors and Gen-IV reactor systems, such as sodium-cooled fast reactors, very high temperature reactors, supercritical water–cooled reactors, and gas-cooled fast reactors. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes ongoing research efforts to deploy new sensors. As described in this paper, a wide range of sensors is available to measure key parameters of interest during fuels and materials irradiations in MTRs. Ongoing development efforts focus on providing MTR users a wider range of parameter measurements with smaller, higher accuracy sensors.

Comparison Measurements of Silicon Carbide Temperature Monitors

As part of a process initiated through the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to make Silicon Carbide (SiC) temperature monitors available for experiments, a capability was developed at the Idaho National Laboratory (INL) to complete post-irradiation evaluations of these monitors. INL selected the resistance measurement approach for detecting peak irradiation temperature from SiC temperature monitors. To demonstrate this new capability, comparison measurements were completed by INL and Oak Ridge National Laboratory (ORNL) on identical samples subjected to identical irradiation conditions. Results reported in this paper indicate that the resistance measurement approach yields similar peak irradiation temperatures if appropriate equipment is used and appropriate procedures are followed.

Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INLs High Temperature Test Laboratory.

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.

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

ATR NSUF Instrumentation Enhancement Efforts

Abstract A key component of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) effort is to expand instrumentation available to users conducting irradiation tests in this unique facility. In particular, development of sensors capable of providing real-time measurements of key irradiation parameters is emphasized because of their potential to increase data fidelity and reduce posttest examination costs. This paper describes the strategy for identifying new instrumentation needed for ATR irradiations and the program underway to develop and evaluate new sensors to address these needs. Accomplishments from this program are illustrated by describing new sensors now available to users of the ATR NSUF. In addition, progress is reported on current research efforts to provide improved in-pile instrumentation to users.

Scaling of Downward Facing Boiling and Steam Venting in a Reactor Vessel/Insulation System

As part of joint U.S.–Korean International Nuclear Engineering Research Initiative (INERI) investigating methods to enhance external cooling of advanced reactor vessel under severe accident conditions, a scaling analysis has been performed to study the phenomena of external cooling of an advanced reactor vessel under severe accident conditions. Five key transfer processes have been considered and the characteristic time for each of these processes has been determined and compared with the residence time for external reactor vessel cooling (ERVC) in the flow channel. To complement the scaling analysis, an ERVC upward co-current two-phase flow model has been developed to predict the total mass flow rate induced in the annular channel by the process of downward facing boiling on the vessel outer surface. The model takes into account the wall heat flux level, the geometry of the vessel/insulation system, the local variation of the cross-sectional flow area, and the pressure drops through various segments of the channel. Based on the results of the ERVC flow calculations and the scaling analysis, criteria for experimental simulation have been established to assure that the ERVC phenomena simulated in laboratory-scale experiments would have the same effects as those anticipated for the full-scale reactor system.Copyright