Insights from Investigations of In-Vessel Retention for High Powered Reactors

In a three-year U.S. - Korean International Nuclear Energy Research Initiative (INERI), state-of-the-art analytical tools and key U.S. and Korean experimental facilities were used to explore two options, enhanced ERVC performance and the use of internal core catchers, that have the potential to increase the margin for in-vessel retention (IVR) in high power reactors (up to 1500 MWe). This increased margin has the potential to improve plant economics (owing to reduced regulatory requirements) and increase public acceptance (owing to reduced plant risk). Although this program focused upon the Korean Advanced Power Reactor -- 1400 MWe (APR 1400) design, recommentations were developed so that they can easily be applied to a wide range of existing and advanced reactor designs. This paper summarizes new data gained for evaluating the margin associated with various options investigated in this program. Insights from analyses completed with this data are also highlighted

Design and Evaluation of an Enhanced In-Vessel Core Catcher

An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (U.S.) - Korean International Nuclear Engineering Research Initiative (INERI) investigating methods to insure In-Vessel Retention (IVR) of core materials that may relocate under severe accident conditions in advanced reactors. 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): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an oxide 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 the status of core catcher design and evaluation efforts, including analyses, materials interaction tests, and prototypic testing efforts

TMI-2 - A Case Study for PWR Instrumentation Performance during a Severe Accident

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated in 2012 by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focussed upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this report. For each sensor, a description is provided with the measured data and conclusions related to the sensor’s survivability, and the basis for conclusions about its survivability. As noted within this document, several techniques were invoked in the TMI-2 post-accident evaluation program to assess sensor status, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this document provides recommendations related to the sensor survivability and data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts

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 advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding of 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 SBLB (Subscale Boundary Layer Boiling) facility at Penn State 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

ATRC Neutron Detector Testing Quick Look Report

As part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program, a joint Idaho State University (ISU) / French Alternative Energies and Atomic Energy Commission (CEA) / Idaho National Laboratory (INL) project was initiated in FY-10 to investigate the feasibility of using neutron sensors to provide online measurements of the neutron flux and fission reaction rate in the ATR Critical Facility (ATRC). A second objective was to provide initial neutron spectrum and flux distribution information for physics modeling and code validation using neutron activation based techniques in ATRC as well as ATR during depressurized operations. Detailed activation spectrometry measurements were made in the flux traps and in selected fuel elements, along with standard fission rate distribution measurements at selected core locations. These measurements provide additional calibration data for the real-time sensors of interest as well as provide benchmark neutronics data that will be useful for the ATR Life Extension Program (LEP) Computational Methods and V&V Upgrade project. As part of this effort, techniques developed by Prof. George Imel will be applied by Idaho State University (ISU) for assessing the performance of various flux detectors to develop detailed procedures for initial and follow-on calibrations of these sensors. In addition to comparing data obtained from each type of detector, calculations will be performed to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. The neutron detectors required for this project were provided to team participants at no cost. Activation detectors (foils and wires) from an existing, well-characterized INL inventory were employed. Furthermore, as part of an on-going ATR NSUF international cooperation, the CEA sent INL three miniature fission chambers (one for detecting fast flux and two for detecting thermal flux) with associated electronics for assessment. In addition, Prof. Imel, ISU, has access to an inventory of Self-Powered Neutron Detectors (SPNDs) with a range of response times as well as Back-to-Back (BTB) fission chambers from prior research he conducted at the Transient REActor Test Facility (TREAT) facility and Neutron RADiography (NRAD) reactors. Finally, SPNDs from the National Atomic Energy Commission of Argentina (CNEA) were provided in connection with the INL effort to upgrade ATR computational methods and V&V protocols that are underway as part of the ATR LEP. Work during fiscal year 2010 (FY10) focussed on design and construction of Experiment Guide Tubes (EGTs) for positioning the flux detectors in the ATRC N-16 locations as well as obtaining ATRC staff concurrence for the detector evaluations. Initial evaluations with CEA researchers were also started in FY10 but were cut short due to reactor reliability issues. Reactor availability issues caused experimental work to be delayed during FY11/12. In FY13, work resumed; and evaluations were completed. The objective of this "Quick Look" report is to summarize experimental activities performed from April 4, 2013 through May 16, 2013

Long Duration Performance of High Temperature Irradiation Resistant Thermocouples

Many advanced nuclear reactor designs require new fuel, cladding, and structural materials. Data are needed to characterize the performance of these new materials in high temperature, radiation conditions. However, traditional methods for measuring temperature inpile degrade at temperatures above 1100 ºC. To address this instrumentation need, the Idaho National Laboratory (INL) developed and evaluated the performance of a high temperature irradiation-resistant thermocouple that contains alloys of molybdenum and niobium. To verify the performance of INL’s recommended thermocouple design, a series of high temperature (from 1200 to 1800 ºC) long duration (up to six months) tests has been initiated. This paper summarizes results from the tests that have been completed. Data are presented from 4000 hour tests conducted at 1200 and 1400 ºC that demonstrate the stability of this thermocouple (less than 2% drift). In addition, post test metallographic examinations are discussed which confirm the compatibility of thermocouple materials throughout these long duration, high temperature tests

Long Duration Testing of Type C Thermocouples at 1500 °C

Experience with Type C thermocouples operating for long periods in the 1400 to 1600 °C temperature range indicate that significant decalibration occurs, often leading to expensive downtime and material waste. As part of an effort to understand the mechanisms causing drift in these thermocouples, the Idaho National Laboratory conducted a long duration test at 1500 °C containing eight Type C thermocouples. As report in this document, results from this long duration test were adversely affected due to oxygen ingress. Nevertheless, results provide key insights about the impact of precipitate formation on thermoelectric response. Post-test examinations indicate that thermocouple signal was not adversely impacted by the precipitates detected after 1,000 hours of heating at 1,500 °C and suggest that the signal would not have been adversely impacted by these precipitates for longer durations (if oxygen ingress had not occurred in this test)

HIGH TEMPERATURE IRRADIATION RESISTANT THERMOCOUPLES – A LOW COST SENSOR FOR IN-PILE TESTING AT HIGH TEMPERATURES

Several options have been identified to improve recently-developed Idaho National Laboratory (INL) High Temperature Irradiation Resistant ThermoCouples (HTIR-TCs) for in-pile testing. These options have the potential to reduce fabrication costs and allow HTIR-TC use in higher temperature applications (up to at least 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 summarizes results from these INL/UI investigations. Specifically, results are reported about several options found to enhance HTIR-TC performance, such as improved heat treatments, alternate geometries, alternate fabrication techniques, and the use of copper/nickel alloys as soft extension cable

In-Pile Thermal Conductivity Measurement Method for Nuclear Fuels

Thermophysical properties of advanced nuclear fuels and materials during irradiation must be known prior to their use in existing, advanced, or next generation reactors. Thermal conductivity is one of the most important properties for predicting fuel and material performance. A joint Utah State University (USU) / Idaho National Laboratory (INL) project, which is being conducted with assistance from the Institute for Energy Technology at the Norway Halden Reactor Project, is investigating in-pile fuel thermal conductivity measurement methods. This paper focuses on one of these methods – a multiple thermocouple method. This two-thermocouple method uses a surrogate fuel rod with Joule heating to simulate volumetric heat generation to gain insights about in-pile detection of thermal conductivity. Preliminary results indicated that this method can measure thermal conductivity over a specific temperature range. This paper reports the thermal conductivity values obtained by this technique and compares these values with thermal property data obtained from standard thermal property measurement techniques available at INL’s High Test Temperature Laboratory. Experimental results and material properties data are also compared to finite element analysis results

DEVELOPMENT OF AN IN-PILE TECHNIQUE FOR THERMAL CONDUCTIVITY MEASUREMENT

Thermophysical properties of advanced fuels and materials during irradiation must be known prior to their use in existing, advanced, or next generation reactors. Fuel thermal conductivity is one of the most important properties for predicting fuel performance and reactor safety. This paper discusses a joint Utah State University (USU)/Idaho National Laboratory (INL) project to investigate an in-pile fuel thermal conductivity measurement technique using a surrogate fuel rod. The method used a surrogate fuel rod with Joule heating to simulate volumetric heat generation as a proof-of-concept test in-pile application. Carbon structural foam, CFOAM®, a product of Touchtone Research Laboratory was chosen as the surrogate material because of the variable electrical and thermal properties upon fabrication. To stay within the surrogate fuel rod requirements, electrical and thermal properties were tailored by Touchtone Research Laboratory to match required values. This paper describes are the techniques used for quantifying thermal conductivity. A description of the test setup and preliminary results are presented. Two thermocouples are inserted into a 1-inch diameter, 6-inch long rod of CFOAM® at known locations. Knowing the applied volumetric heat to the rod by electrical resistance heating, the thermal conductivity can be calculated. Sensitivities of this measurement can also found by analysis and testing of different configurations of the sample setup. Verification of thermal conductivity is found by measuring the thermal properties of the CFOAM® using different methods. Thermal properties including thermal conductivity, specific heat capacity, and expansion coefficient of two types of CFOAM®, CFOAM20 and CFOAM25, were characterized using standard measurement techniques, such as laser flash, differential scanning calorimetry, and pushrod dilatometry