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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85 Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Advancements in FeCrAl Alloys for Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors

Advanced FeCrAl alloys are being considered as an enhanced accident tolerant fuel (ATF) cladding for light water reactor (LWR) applications as they can exhibit slower reaction kinetics with high temperature steam and have a reduced production of hydrogen during accident scenarios compared both to historic 304L cladding and Zircaloy. Both of these attributes are key towards producing a fuel-form with enhanced accident tolerance over the current Zr-based-UO{sub 2} fuel form in deployment. A multi-year, multi-laboratory effort has been established to rapidly advance FeCrAl alloys as an ATF concept. This effort has focused on optimizing the microstructure and composition of these alloys in an attempt to provide performance on-par with or better than the current Zr-based cladding in both normal operating conditions and under accident scenarios. To date, a significant database supporting key performance factors in both scenarios has been developed. Based on this database, a compositional and microstructural window has been identified providing the insight for an optimized, advanced nuclear-grade FeCrAl alloy for LWRs. This work will provide a comprehensive overview of the systematic effort to advance FeCrAl alloys as an enhanced ATF cladding in commercial LWRs. Efforts undertaken by this multi-year, multi-laboratory program have shown the viability of using FeCrAl alloys as an ATF cladding in LWRs. Most importantly, it has been shown the oxidation resistance, decreased hydrogen production, and reduced heat generation of FeCrAl alloys strongly contribute to its ATF capability. Furthermore, it has been shown that the normal operational performance of a FeCrAl cladding can rival that of Zr-based alloys. Finally, it was found that optimizing the performance of FeCrAl cladding materials requires balancing the starting composition and microstructure in order to satisfy key performance indicators such as oxidation resistance and radiation tolerance. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set of multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the individual coupling coefficients from measurement: unlike the symmetrical cases, only the product of the values can be obtained. A method is proposed utilizing MCNP6 tally ratios to separate the coupling coefficients for such systems. This work provides insight into the behavior of asymmetrically-coupled systems as the separation distance between the two cores is changed and also as the asymmetry is increased. As the asymmetry increases, both the slower and the faster observable prompt neutron decay constants increase in magnitude. The coupling time constants are determined from the measured decay constants. As the separation distance increases, both coupling coefficients decrease as expected. Based on these findings, an effective computational method utilizing MCNP6 and the Rossialpha technique can be applied to the prediction of asymmetrical coupled system measurements.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Toward a clinical application of ex situ boron neutron capture therapy for lung tumors at the RA-3 reactor in Argentina

Purpose: Many types of lung tumors have a very poor prognosis due to their spread in the whole organ volume. The fact that boron neutron capture therapy (BNCT) would allow for selective targeting of all the nodules regardless of their position, prompted a preclinical feasibility study of ex situ BNCT at the thermal neutron facility of RA-3 reactor in the province of Buenos Aires, Argentina. (L)-4p-dihydroxy-borylphenylalanine fructose complex (BPA-F) biodistribution studies in an adult sheep model and computational dosimetry for a human explanted lung were performed to evaluate the feasibility and the therapeutic potential of ex situ BNCT. Methods: Two kinds of boron biodistribution studies were carried out in the healthy sheep: a set of pharmacokinetic studies without lung excision, and a set that consisted of evaluation of boron concentration in the explanted and perfused lung. In order to assess the feasibility of the clinical application of ex situ BNCT at RA-3, a case of multiple lung metastases was analyzed. A detailed computational representation of the geometry of the lung was built based on a real collapsed human lung. Dosimetric calculations and dose limiting considerations were based on the experimental results from the adult sheep, and on the most suitable information published in the literature. In addition, a workable treatment plan was considered to assess the clinical application in a realistic scenario. Results: Concentration-time profiles for the normal sheep showed that the boron kinetics in blood, lung, and skin would adequately represent the boron behavior and absolute uptake expected in human tissues. Results strongly suggest that the distribution of the boron compound is spatially homogeneous in the lung. A constant lung-to-blood ratio of 1.3 ± 0.1 was observed from 80 min after the end of BPA-F infusion. The fact that this ratio remains constant during time would allow the blood boron concentration to be used as a surrogate and indirect quantification of the estimated value in the explanted healthy lung. The proposed preclinical animal model allowed for the study of the explanted lung. As expected, the boron concentration values fell as a result of the application of the preservation protocol required to preserve the lung function. The distribution of the boron concentration retention factor was obtained for healthy lung, with a mean value of 0.46 ± 0.14 consistent with that reported for metastatic colon carcinoma model in rat perfused lung. Considering the human lung model and suitable tumor control probability for lung cancer, a promising average fraction of controlled lesions higher than 85% was obtained even for a low tumor-to-normal boron concentration ratio of 2. Conclusions: This work reports for the first time data supporting the validity of the ovine model as an adequate human surrogate in terms of boron kinetics and uptake in clinically relevant tissues. Collectively, the results and analysis presented would strongly suggest that ex situ whole lung BNCT irradiation is a feasible and highly promising technique that could greatly contribute to the treatment of metastatic lung disease in those patients without extrapulmonary spread, increasing not only the expected overall survival but also the resulting quality of life.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

ORNL Evaluation of Electrabel Safety Cases for Doel 3 / Tihange 2: Final Report

Oak Ridge National Laboratory (ORNL) performed a detailed technical review of the 2015 Electrabel (EBL) Safety Cases prepared for the Belgium reactor pressure vessels (RPVs) at Doel 3 and Tihange 2 (D3/T2). The Federal Agency for Nuclear Control (FANC) in Belgium commissioned ORNL to provide a thorough assessment of the existing safety margins against cracking of the RPVs due to the presence of almost laminar flaws found in each RPV. Initial efforts focused on surveying relevant literature that provided necessary background knowledge on the issues related to the quasilaminar flaws observed in D3/T2 reactors. Next, ORNL proceeded to develop an independent quantitative assessment of the entire flaw population in the two Belgian reactors according to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Appendix G, Fracture Toughness Criteria for Protection Against Failure, New York (1992 and 2004). That screening assessment of all EBL-characterized flaws in D3/T2 used ORNL tools, methodologies, and the ASME Code Case N-848, Alternative Characterization Rules for QuasiLaminar Flaws . Results and conclusions from the ORNL flaw acceptance assessments of D3/T2 were compared with those from the 2015 EBL Safety Cases. Specific findings of the ORNL evaluation of that part of the EBL structural integrity assessment focusing on stability of the flaw population subjected to primary design transients include the following: ORNL s analysis results were similar to those of EBL in that very few characterized flaws were found not compliant with the ASME (1992) acceptance criterion. ORNL s application of the more recent ASME Section XI (2004) produced only four noncompliant flaws, all due to LOCAs. The finding of a greater number of non-compliant flaws in the EBL screening assessment is due principally to a significantly more restrictive (conservative) criterion for flaw size acceptance used by EBL. ORNL s screening assessment results (obtained using an analysis methodology different from that of EBL) are interpreted herein as confirming the EBL screening results for D3/T2. ORNL s independent refined analysis demonstrated the EBL-characterized flaw 1660, which is non-compliant in the ORNL and EBL screening assessment, is rendered compliant when modeled as a more realistic individual quasi-laminar flaw using a 3-D XFEM analysis approach. ORNL s and EBL s refined analyses are in good agreement for the flaw 1660 close to the clad/base metal interface; ORNL is not persuaded that repeating this exercise for more than one non-compliant flaw is necessary to accept the EBL conclusions derived from the aggregate of EBL refined analysis results. ORNL General Conclusions Regarding the Structural Integrity Assessment (SIA) Conducted by EBL for D3/T2 Based on comparative evaluations of ORNL and EBL SIA analyses and on consideration of other results, ORNL is in agreement with the general conclusions reported by Electrabel in their RPV D3/T2 Technical Summary Note of April 14, 2015: More than 99 percent of flaws in D3/T2 meet the defined screening criterion, rendering them benign with respect to initiation in the event of a design transient. Refined analyses of non-compliant flaws from the screening assessment indicate that only 11 of the 16196 detected flaws have a critical reference-temperature material index (designated RTNDT) that implies the possibility of the initiation of cleavage fracture at some future time. For those 11 2 flaws, the calculated margin in RTNDT (a measure of acceptable embrittlement relative to end-ofservice-life conditions) is significant, being greater than 80 C. Fatigue crack growth is not a concern in the flaw-acceptability analyses. Primary stress re-evaluation confirms that the collapse pressure is more than 1.5 times the design pressure in the presence of defects detected in D3/T2. Sufficient conservatisms are built into the input data and into the different steps of the SIA; in some cases, those conservatisms are quantified and imply that additional margins exist in the SIA. Taken as a whole, the foregoing results and conclusions confirm the structural integrity of Doel 3 and Tihange 2 under all design transients with ample margin in the presence of the 16196 detected flaws.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Research to Develop Flood Barrier Testing Strategies for Nuclear Power Plants

The U.S. Nuclear Regulatory Commission has developed regulations regarding the siting and design of nuclear power plants (NPPs) that are aimed at addressing various natural hazards, including flooding. Flood barriers are designed to prevent water from entering NPP areas containing structures, systems, and components (SSCs) important to safety. The barriers are used at NPPs along with drains, sumps, pumps, valves, plugs, and site grading as part of the plant flood protection features that protect SSCs from experiencing external or internal flooding and mitigate the effects of flooding on NPP operations. The performance of flood protection features, including flood barriers at NPPs, has been an ongoing concern. Domestic and international operational experience provides clear indications that flood barrier performance has significant safety implications, especially for aging NPPs. The observed deficiencies show that flood barriers should be designed and installed properly, then adequately tested, inspected, and maintained in order to ensure that they perform their intended functions during flooding events. Here, this paper reviews available information related to flood barriers employed at U.S. NPPs and provides an overview and categorization of NPP flood barriers. It identifies potential domestic and international flood barrier testing facilities, including operating and decommissioned U.S. NPPs. Finally, this paper presents the technical and logistical considerations that should be made when developing specific testing strategies and protocols for flood barriers, such as the selection of flood barriers, test locations, testing approach, performance criteria, and testing parameters.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America

The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Completion of Level 4 Milestone M4AT-15OR2301039 for the Johnson Noise Thermometry for Drift-free Temperature Measurements Work Package AT-15OR230103

This memorandum constitutes our September 2015 level 4 milestone for the project entitled “Johnson Noise Thermometry for Drift-free Temperature Measurements” and satisfies the Milestone/Activity (Conclude HFIR field demonstration of JNT prototype). The progress summary describes the work performed to complete the subject milestone.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Computer Security for Commercial Nuclear Power Plants - Literature Review for Korea Hydro Nuclear Power Central Research Institute

Sandia National Laboratories (SNL) is providing training and consultation activities on security planning and design for the Korea Hydro and Nuclear Power Central Research Institute (KHNPCRI). As part of this effort, SNL performed a literature review on computer security requirements, guidance and best practices that are applicable to an advanced nuclear power plant. This report documents the review of reports generated by SNL and other organizations [U.S. Nuclear Regulatory Commission, Nuclear Energy Institute, and International Atomic Energy Agency] related to protection of information technology resources, primarily digital controls and computer resources and their data networks. Copies of the key documents have also been provided to KHNP-CRI.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

MELCOR/CONTAIN LMR Implementation Report - FY16 Progress.

This report describes the progress of the CONTAIN - LMR sodium physics and chemistry models to be implemented in MELCOR 2.1. In the past three years , the implementation included the addition of sodium equations of state and sodium properties from two different sources. The first source is based on the previous work done by Idaho National Laboratory by modifying MELCOR to include liquid lithium equation of state as a working fluid to model the nuclear fusion safety research. The second source uses properties generated for the SIMMER code. The implemented modeling has been tested and results are reported in this document. In addition, the CONTAIN - LMR code was derived from an early version of the CONTAIN code, and many physical models that were developed since this early version of CONTAIN are not available in this early code version. Therefore, CONTAIN 2 has been updated with the sodium models in CONTAIN - LMR as CONTAIN2 - LMR, which may be used to provide code-to-code comparison with CONTAIN - LMR and MELCOR when the sodium chemistry models from CONTAIN - LMR have been completed. Both the spray fire and pool fire chemistry routines from CONTAIN - LMR have been integrated into MELCOR 2.1, and debugging and testing are in progress. Because MELCOR only models the equation of state for liquid and gas phases of the coolant, a modeling gap still exists when dealing with experiments or accident conditions that take place when the ambient temperature is below the freezing point of sodium. An alternative method is under investigation to overcome this gap . We are no longer working on the separate branch from the main branch of MELCOR 2.1 since the major modeling of MELCOR 2.1 has been completed. At the current stage, the newly implemented sodium chemistry models will be a part of the main MELCOR release version (MELCOR 2.2). This report will discuss the accomplishments and issues relating to the implementation. Also, we will report on the planned completion of all remaining tasks in the upcoming FY2017, including the atmospheric chemistry model and sodium - concrete interaction model implementation .

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

L3:PHI.CMD.P13.02 Support for CILC L1 Milestone Using STAR-CCM+

This report documents work performed to support Consortium for the Advanced Simulation of LWRs (CASL) modeling of Chalk River Unidentified Deposit (CRUD) Induced Power Shift (CIPS) and CRUD Induced Local Corrosion (CILC) using the Cicada package. The work documented here is intended to complement current and future CIPS and CILC modeling activities in CASL. We provide tools for crud and corrosion-related simulation and analysis by developing a better understanding of the interplay between the coupled physics that describe the phenomena at different time and length scales. We intend to use these models to better inform future simulation capability and development.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Headspace analysis of foams and fixatives

In the process of decontaminating and decommissioning (D&D) older nuclear facilities, special precautions must be taken with removable or airborne contamination. One possible strategy utilizes foams and fixatives to affix these loose contaminants. Many foams and fixatives are already commercially available, either generically or sold specifically for D&D. However, due to a lack of revelant testing in a radioactive environment, additional verification is needed to confirm that these products not only affix contamination to their surfaces, but also will function in a D&D environment. Several significant safety factors, including flammability and worker safety, can be analyzed through the process of headspace analysis, a technique that analyzes the off gas formed before or during the curing process of the foam/fixative, usually using gas chromatography-mass spectrometry (GC-MS). This process focuses on the volatile components of a chemical, which move freely between the solid/liquid form within the sample and the gaseous form in the area above the sample (the headspace). Between possibly hot conditions in a D&D situation and heat created in a foaming reaction, the volatility of many chemicals can change, and thus different gasses can be released at different times throughout the reaction. This project focused on analysis of volatile chemicals involved in the process of using foams and fixatives to identify any potential hazardous or flammable compounds.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Thermochemical and transport properties important to molten salt reactor operation: Off-gas performance and the fission product mechanistic source term

The source term used in reactor performance and safety analysis requires understanding the chemical speciation and transport behavior of fission and activation products in the primary heat transport system (PHTS) and containment. As a starting condition, the transport models will use the isotopic compositions determined through ORIGEN (Bell 1973) calculations and thermochemistry that describes the speciation of the fuel, the salt, activation products and fission products. However, as changes in phase will be under non-equilibrium conditions, the source terms will also need models for the transport behavior of key compounds in the PHTS under various scenarios: normal operation with and without an off-gas system, maintenance, off-normal events, and severe reactor accidents. This document highlights those mechanistic behaviors for which models are needed. Simulations of reactor behavior require these models to be developed for molten salt reactors (MSRs), especially if not part of the suite used in describing LWR behavior. The document concludes with recommendations how to develop these models based on what is already known about reactor safety and fission product behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Initial development and extension of EPP methods to Grade 91

This report describes modifications to the existing elastic perfectly plastic (EPP) analysis ASME Nuclear Code Cases N-861 and N-862 required to apply the Code Case design methods to Grade 91 steel. The critical difference between the existing Code Cases, applicable only to 304H and 316H stainless steel, is that they do not account for material cyclic softening. Grade 91 is a softening material at elevated temperature. As such, modifications to the existing Code Case rules are required before they can be conservatively applied to Grade 91.This report describes modified EPP methods applicable to softening materials like Grade 91. Additionally, the report describes the specific technical and administrative changes to the existing EPP Code Cases that will be required to support Grade 91.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Report on the FY20 Characterization of Age Hardened Alloy 709 Materials

This report provides the results on microstructural characterization of the first commercial heat of Alloy 709 plate material undergone additional age hardening treatment. In FY19, through thermodynamic-kinetic modeling and microstructural characterization, it is determined that the 775°C/10 h age-hardening treatment can promote desired precipitation on dislocations and has the potential to further improve the mechanical performance of the material. In FY20, the treatment protocol was applied to the ESR-1150°C condition, and the microstructure was then compared with the ESR-1100°C condition. It was found that the 1150°C solution annealing (SA) dramatically reduced the dislocation density compared to the 1100°C SA, resulting a smaller number density of Nb-rich fine precipitates which grow on dislocations during the age hardening treatment. Other features are very similar between the two. The effect of age hardening on creep performance is discussed. In addition, TEM characterization of a creep-ruptured sample (600°C/330 MPa) was performed. Specimens from the grip region and the gauge region was studied. The grip region specimen was effectively aged at 600°C for extended period, while the gauge region specimen was mechanically deformed at 600°C. The microstructures are very different, implying the importance of a dynamical effect on precipitation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Post-Fukushima Research and Development Strategy for MELCOR

Numerous MELCOR modeling improvements and analyses have been performed in the time since the severe accidents at Fukushima Daiichi Nuclear Power Station that occurred in March 2011. This report briefly summarizes the related accident reconstruction and uncertainty analysis efforts. It further discusses a number of potential pursuits to further advance MELCOR modeling and analysis of the severe accidents at Fukushima Daiichi and severe accident modeling in general. Proposed paths forward include further enhancements to identified MELCOR models primarily impacting core degradation calculations, and continued application of uncertainty analysis methods to improve model performance and a develop deeper understanding of severe accident progression.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗