Publications

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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.

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In response to the expansion of nuclear fuel cycle (NFC) activities (and the associated suite of risks) around the world, this effort provides an evaluation of systems-based solutions for managing such risk complexity in multi-modal (land and water), and multi-jurisdictional international spent nuclear fuel (SNF) transportation. By better understanding systemic risks in SNF transportation, developing SNF transportation risk assessment frameworks, and evaluating these systems-based risk assessment frameworks, this research illustrates interdependency between safety, security, and safeguards (3S) risks is inherent in NFC activities that can go unidentified when each “S” is independently evaluated. Two novel system-theoretic analysis techniques, dynamic probabilistic risk assessment (DPRA) and system-theoretic process analysis (STPA), provide integrated 3S analysis to address these interdependencies. This research suggests a need (and provides a way) to reprioritize United States engagement efforts to reduce global SNF transportation risks. Note: This paper is a summary of the final results found in Reference [1].

The U.S. Nuclear Regulatory Commission initiated the state-of-the-art reactor consequence analyses (SOARCA) project to develop realistic estimates of the offsite radiological health consequences for potential severe reactor accidents. The SOARCA analysis of an ice condenser containment plant was performed because its relatively low design pressure and reliance on igniters makes it potentially susceptible to early containment failure from hydrogen combustion during a severe accident. The focus was on station blackout accident scenarios where all alternating current power is lost. Accident progression calculations used the MELCOR computer code and offsite consequence analyses were performed with MACCS. The analysis included more than 500 MELCOR and MACCS simulations to account for uncertainty in important accident progression and offsite consequence input parameters. Consequences from severe nuclear power plant accidents modeled in this and previous SOARCA analyses are smaller than calculated in earlier studies. The delayed releases calculated provide more time for emergency response actions. The results show that early containment failure is very unlikely, even without successful use of igniters. However, these results are dependent on the distributions assigned to safety valve failure-to-close parameters, and considerable uncertainty remains on the true distributions for these parameters due to very limited test data. Even for scenarios resulting in early containment failure, the calculated individual latent fatal cancer risks are very small. Early and latent-cancer fatality risks are one focus of this paper. Regression results showing the most influential parameters are also discussed.

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To support more rigorous analysis on global security issues at Sandia National Laboratories (SNL), there is a need to develop realistic data sets without using "real" data or identifying "real" vulnerabilities, hazards or geopolitically embarrassing shortcomings. In response, an interdisciplinary team led by subject matter experts in SNL's Center for Global Security and Cooperation (CGSC) developed a hypothetical case description. This hypothetical case description assigns various attributes related to international SNF transportation that are representative, illustrative and indicative of "real" characteristics of "real" countries. There is no intent to identify any particular country and any similarity with specific real-world events is purely coincidental. To support the goal of this report to provide a case description (and set of scenarios of concern) for international SNF transportation inclusive of as much "real-world" complexity as possible -- without crossing over into politically sensitive or classified information -- this SAND report provides a subject matter expert-validated (and detailed) description of both technical and political influences on the international transportation of spent nuclear fuel.

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The purpose of the scenarios workshop held for the Civilian Nuclear component of the Global Nuclear Assured Security Mission Integration Initiative was to identify sources of risk in the global civilian nuclear enterprise. The risks identified are inadequately addressed through current technical measures, regulatory frameworks and institutions and should be considered for further research. The workshop participants also developed four high level scenarios describing different sequences of events that could result in radiological releases, widespread loss of electric power, and loss of public confidence in segments of the nuclear industry. The scenarios are intended for further analysis and as the basis for simulation exercises.

In response to the expansion of nuclear fuel cycle (NFC) activities -- and the associated suite of risks -- around the world, this project evaluated systems-based solutions for managing such risk complexity in multimodal and multi-jurisdictional international spent nuclear fuel (SNF) transportation. By better understanding systemic risks in SNF transportation, developing SNF transportation risk assessment frameworks, and evaluating these systems-based risk assessment frameworks, this research illustrated interdependency between safety, security, and safeguards risks is inherent in NFC activities and can go unidentified when each "S" is independently evaluated. Two novel system-theoretic analysis techniques -- dynamic probabilistic risk assessment (DPRA) and system-theoretic process analysis (STPA) -- provide integrated "3S" analysis to address these interdependencies and the research results suggest a need -- and provide a way -- to reprioritize United States engagement efforts to reduce global nuclear risks. Lastly, this research identifies areas where Sandia National Laboratories can spearhead technical advances to reduce global nuclear dangers.

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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.

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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).

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.

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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.

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