A key objective of the United States Department of Energy’s (DOE) Office of Nuclear Energy’s Spent Fuel and Waste Science and Technology Campaign is to better understand the technical basis, risks, and uncertainty associated with the safe and secure disposition of spent nuclear fuel (SNF) and high-level radioactive waste. Commercial nuclear power generation in the United States has resulted in thousands of metric tons of SNF, the disposal of which is the responsibility of the DOE (Nuclear Waste Policy Act of 1982, as amended). Any repository licensed to dispose of SNF must meet requirements regarding the long-term performance of that repository. For an evaluation of the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration during the postclosure period. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently licensed and being used to store and transport SNF but were not designed for permanent geologic disposal. A study has been initiated to examine the potential consequences, with respect to long-term repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. The first phase (a scoping phase) consisted of developing an approach to creating the modeling tools and techniques that may eventually be needed to either include or exclude criticality from a performance assessment (PA) as appropriate; this scoping phase is documented in Price et al. (2019a). In the second phase, that modeling approach was implemented and future work was identified, as documented in Price et al. (2019b). This report gives the results of a repository-scale PA examining the potential consequences of postclosure criticality, as well as the information, modeling tools, and techniques needed to incorporate the effects of postclosure criticality in the PA.
Coupling interests in small modular reactors (SMR) as efficient and effective method to meet increasing energy demands with a growing aversion to cost and schedule overruns traditionally associated with the current fleet of commercial nuclear power plants (NPP), SMRs are attractive because they offer a significant relative cost reduction to current-generation nuclear reactors-- increasing their appeal around the globe. Sandia's Global Nuclear Assurance and Security (GNAS) research perspective reframes the discussion around the "complex risk" of SMRs to address interdependencies between safety, safeguards, and security. This systems study provides technically rigorous analysis of the safety, safeguards, and security risks of SMR technologies. The aims of this research is three-fold. The first aim is to provide analytical evidence to support safety, safeguards, and security claims related to SMRs (Study Report Volume I). Second, this study aims to introduce a systems-theoretic approach for exploring interdependencies between the technical evaluations (Study Report Volume II). The third aim is to demonstrate Sandia's capability for timely, rigorous, and technical analysis to support emerging complex GNAS mission objectives. This page left blank intentionally
This document details the Fiscal Year 2016 modeling efforts to define the true operating limitations (margins) of the Terry turbopump systems used in the nuclear industry for Milestone 3 (full-scale component experiments) and Milestone 4 (Terry turbopump basic science experiments) experiments. The overall multinational-sponsored program creates the technical basis to: (1) reduce and defer additional utility costs, (2) simplify plant operations, and (3) provide a better understanding of the true margin which could reduce overall risk of operations.
This document details the milestone approach to define the true operating limitations (margins) of the Terry turbopump systems used in the nuclear industry for Milestone 3 (full-scale component experiments) and Milestone 4 (Terry turbopump basic science experiments) efforts. The overall multinational-sponsored program creates the technical basis to: (1) reduce and defer additional utility costs, (2) simplify plant operations, and (3) provide a better understanding of the true margin which could reduce overall risk of operations.
The Terry turbine is a small, single-stage, compound-velocity impulse turbine originally designed and manufactured by the Terry Steam Turbine Company purchased by Ingersoll-Rand in 1974. Terry turbines are currently manufactured and marketed by Dresser-Rand. Terry turbines were principally designed for waste-steam applications. Terry turbopumps are ubiquitous to the US nuclear fleet as a steam driven turbopump in either the reactor core isolation cooling system (RCIC) and high pressure coolant injection systems for boiling water reactors (BWRs) or in the auxiliary feedwater system (AFW) system for pressurized water reactors (PWRs).
ANS IHLRWM 2017 - 16th International High-Level Radioactive Waste Management Conference: Creating a Safe and Secure Energy Future for Generations to Come - Driving Toward Long-Term Storage and Disposal
Transportation of spent nuclear fuel (SNF) is expected to increase in the future, as the nuclear fuel infrastructure continues to expand and fuel takeback programs increase in popularity. Analysis of potential risks and threats to SNF shipments is currently performed separately for safety and security. However, as SNF transportation increases, the plausible threats beyond individual categories and the interactions between them become more apparent. A new approach is being developed to integrate safety, security, and safeguards (3S) under a system-theoretic framework and a probabilistic risk framework. At the first stage, a simplified scenario will be implemented using a dynamic probabilistic risk assessment (DPRA) method. This scenario considers a rail derailment followed by an attack. The consequences of derailment are calculated with RADTRAN, a transportation risk analysis code. The attack scenarios are analyzed with STAGE, a combat simulation model. The consequences of the attack are then calculated with RADTRAN. Note that both accident and attack result in SNF cask damage and a potential release of some fraction of the SNF inventory into the environment. The major purpose of this analysis was to develop the input data for DPRA. Generic PWR and BWR transportation casks were considered. These data were then used to demonstrate the consequences of hypothetical accidents in which the radioactive materials were released into the environment. The SNF inventory is one of the most important inputs into the analysis. Several pressurized water reactor (PWR) and boiling water reactor (BWR) fuel burnups and discharge times were considered for this proof-of-concept. The inventory was calculated using ORIGEN (point depletion and decay computer code, Oak Ridge National Laboratory) for 3 characteristic burnup values (40, 50, and 60 GWD/MTU) and 4 fuel ages (5, 10, 25 and 50 years after discharge). The major consequences unique to the transportation of SNF for both accident and attack are the results of the dispersion of radionuclides in the environment. The dynamic atmospheric dispersion model in RADTRAN was used to calculate these consequences. The examples of maximum exposed individual (MEI) dose, early mortality and soil contamination are discussed to demonstrate the importance of different factors. At the next stage, the RADTRAN outputs will be converted into a form compatible with the STAGE analysis. As a result, identification of additional risks related to the interaction between characteristics becomes a more straightforward task. In order to present the results of RADTRAN analysis in a framework compatible with the results of the STAGE analysis, the results will be grouped into three categories: • Immediate negative harms •Future benefits that cannot be realized •Additional increases in future risk By describing results within generically applicable categories, the results of safety analysis are able to be placed in context with the risk arising from security events.
Radionuclide inventories are generated to permit detailed analyses of the Fukushima Daiichi meltdowns. This is necessary information for severe accident calculations, dose calculations, and source term and consequence analyses. Inventories are calculated using SCALE6 and compared to values predicted by international researchers supporting the OECD/NEA's Benchmark Study on the Accident at Fukushima Daiichi Nuclear Power Station (BSAF). Both sets of inventory information are acceptable for best-estimate analyses of the Fukushima reactors. Consistent nuclear information for severe accident codes, including radionuclide class masses and core decay powers, are also derived from the SCALE6 analyses. Key nuclide activity ratios are calculated as functions of burnup and nuclear data in order to explore the utility for nuclear forensics and support future decommissioning efforts.
Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine decreases the developed turbine torque; the RCIC speed then slows, and thus the pump flow rate to the RPV decreases. Subsequently, RPV water level decreases due to continued boiling and the liquid fraction flowing to the RCIC decreases, thereby accelerating the RCIC and refilling the RPV. The feedback cycle then repeats itself and/or reaches a quasi-steady equilibrium condition. In other words, the water carry-over is limited by cyclic RCIC performance degradation, and hence the system becomes self-regulating. The indications achieved to date with the system model are more qualitative than quantitative. The avenues being pursued to increase the fidelity of the model are expected to add quantitative realism. The end product will be generic in the sense that the RCIC model will be incorporable within the larger reactor coolant system model of any nuclear power plant or experimental configuration.