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...

This document presents the results from a U.S. Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various :_evere accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-fc_rm or simplified numerical solution techniques. Finite element techniques were employed for analytical model verification and examining more detailed phcnomena.High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

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SUMMARYNew and enhanced nuclear fuels are a key enabler for new and improved reactor technologies. For example, the goals of the next generation nuclear plant (NGNP) will not be met without irradiations successfully demonstrating the safety and reliability of new fuels. Likewise, fuel reliability has become paramount in ensuring the competitiveness of nuclear power plants. Recently, the Office of Nuclear Energy in the Department of Energy (DOE-NE) launched a new direction in fuel research and development that emphasizes an approach relying on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time, data are essential for characterizing the performance of new fuels during irradiation testing. A three-year strategic research program has been initiated for developing the required test vehicles with sensors of unprecedented accuracy and resolution for obtaining the data needed to characterize three-dimensional changes in fuel microstructure during irradiation testing. When implemented, this strategy will yield test capsule designs that are instrumented with new sensor technologies for irradiations at facilities primarily relied upon by the Fuel Cycle Research and Development (FCR&D) program, the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR). Prior laboratory testing, and as needed, irradiation testing of sensors in these capsules will have been completed to give sufficient confidence that the irradiation tests will yield the required data.From the onset of this instrumentation development effort, it was recognized that obtaining these sensors must draw upon the expertise of a wide-range of organizations not currently supporting nuclear fuels research. Hence, a draft version of this document was developed to provide necessary background information related to fuel irradiation testing, desired parameters for detection, and an overview of currently available in-pile instrumentation. Then, a workshop was held in which U.S. and foreign experts from fuels, irradiation, and instrumentation fields participated. Prior to this workshop, copies of a draft version of this document were distributed to participants to stimulate expert interactions at this meeting. During the workshop, candidate sensor technologies identified in this document were discussed and ranked by the experts using agreed upon criteria. The final version of this document describes the consensus reached during the workshop with respect to recommendations for the path forward for accomplishing the goals of this research program.Based on the activities completed to develop this strategic plan, it is recommended that the FCR&D instrumentation development program be initiated as a three year program that includes the following three tasks:• Ultrasonics-Based Evaluations -In this task, laboratory evaluations and necessary irradiations will be completed to demonstrate the viability of this technology for in-pile applications. Specif...

Ultrasonic thermometry sensors (UTS) have been intensively studied in the past to measure temperatures from 2080 K to 3380 K. This sensor, which uses the temperature dependence of the acoustic velocity in materials, was developed for experiments in extreme environments. Its major advantages, which are (a) capability of measuring a temperature profile from multiple sensors on a single probe and (b) measurement near the sensor material melting point, can be of great interest when dealing with on-line monitoring of high-temperature safety tests. Ultrasonic techniques were successfully applied in several severe accident related experiments. With new developments of alternative materials, this instrument may be used in a wide range of experimental areas where robustness and compactness are required. Long-term irradiation experiments of nuclear fuel to extremely high burn-ups could benefit from this previous experience. After an overview of UTS technology, this article summarizes experimental work performed to improve the reliability of these sensors. The various C. Wilkins is working as a consultant (retired from INL). designs, advantages, and drawbacks are outlined and future prospects for long-term high-temperature irradiation experiments are discussed.

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 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 focused 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 paper. As noted within this paper, several techniques were invoked in the TMI-2 post-accident program to evaluate sensor survivability status and data qualification, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in largescale 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 paper provides recommendations related to sensor survivability and the data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

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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