Randall O. Gauntt

Fukushima Daiichi accident study : status as of April 2012.

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission (NRC) and Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing severe accident modeling capability of the MELCOR code. MELCOR is the state-of-the-art system-level severe accident analysis code used by the NRC to provide information for its decision-making process in this area. The objectives of the project were: (1) collect, verify, and document data on the accidents by developing an information portal system; (2) reconstruct the accident progressions using computer models and accident data; and (3) validate the MELCOR code and the Fukushima models, and suggest potential future data needs. Idaho National Laboratory (INL) developed an information portal for the Fukushima Daiichi accident information. Sandia National Laboratories (SNL) developed MELCOR 2.1 models of the Fukushima Daiichi Units 1, 2, and 3 reactors and the Unit 4 spent fuel pool. Oak Ridge National Laboratory (ORNL) developed a MELCOR 1.8.5 model of the Unit 3 reactor and a TRACE model of the Unit 4 spent fuel pool. The good correlation of the results from the SNL models with the data from the plants and with the ORNL model results provides additional confidence in the MELCOR code. The modeling effort has also provided insights into future data needs for both model development and validation.

Transient Analysis of Sulfur-Iodine Cycle Experiments and Very High Temperature Reactor Simulations Using MELCOR-H2

Abstract MELCOR is a thermal-hydraulic code used by the United States and the international nuclear community for the modeling of both light water and gas-cooled reactors. MELCOR was extended in order to model nuclear reactors that are coupled to the sulfur-iodine (SI) cycle for cogeneration of hydrogen. This version of the code is known as MELCOR-H2, and it includes modular secondary system components (e.g., turbines, compressors, heat exchangers, and generators), a point-kinetics model, and a graphical user interface. MELCOR-H2 allows for the fully coupled, transient analysis and design of the nuclear thermochemical SI cycle for the purpose of maximizing the production of hydrogen and electricity. Recent work has demonstrated that the hydrogen generation rate calculated by MELCOR-H2 for the SI cycle was within the expected theoretical yield. In order to benchmark MELCOR-H2, we simulated a set of sulfuric acid decomposition experiments that were conducted at Sandia National Laboratories during 2006. We also used MELCOR-H2 to simulate a 2004 Japan Atomic Energy Research Institute SI experiment. The simulations compared favorably with both experiments; most measured and calculated outputs were within 10%. The simulations adequately calculated O2, SO2, and H2 production rate, acid conversion efficiency, the relationship between solution mole percent and conversion efficiency, and the relationship between molar flow rate and efficiency. We also simulated a 6-stage turbine and a 20-stage compressor. Our results were mostly within 1 or 2% of the literature. Then, we simulated a pebble bed very high temperature reactor (VHTR) and compared key MELCOR-H2 results with the literature. The comparison showed that the results were typically within 1 or 2%. Finally, we compared the MELCOR-H2 point-kinetics model with the exact Inhour reactivity solution for various cases, including a 1.0

Revisiting Insights from Three Mile Island Unit 2 Postaccident Examinations and Evaluations in View of the Fukushima Daiichi Accident

step reactivity insertion. We were able to employ a large time step while successfully matching the theoretical power level. These comparisons demonstrate MELCOR-H2’s unique ability to simulate fully coupled VHTRs for the production of hydrogen.

MELCOR Simulations of the Severe Accident at Fukushima Daiichi Unit 3

Abstract The Three Mile Island Unit 2 (TMI-2) accident, which occurred on March 28, 1979, led industry and regulators to enhance strategies to protect against severe accidents in commercial nuclear power plants. Investigations in the years after the accident concluded that at least 45% of the core had melted and that nearly 19 tonnes of the core material had relocated to the lower head. Postaccident examinations indicate that about half of that material formed a solid layer near the lower head and above it was a layer of fragmented rubble. As discussed in this paper, numerous insights related to pressurized water reactor accident progression were gained from postaccident evaluations of debris, reactor pressure vessel (RPV) specimens, and nozzles taken from the RPV. In addition, information gleaned from TMI-2 specimen evaluations and available data from plant instrumentation were used to improve severe accident simulation models that form the technical basis for reactor safety evaluations. Finally, the TMI-2 accident led the nuclear community to dedicate considerable effort toward understanding severe accident phenomenology as well as the potential for containment failure. Because available data suggest that significant amounts of fuel heated to temperatures near melting, the events at Fukushima Daiichi Units 1, 2, and 3 offer an unexpected opportunity to gain similar understanding about boiling water reactor accident progression. To increase the international benefit from such an endeavor, we recommend that an international effort be initiated to (a) prioritize data needs; (b) identify techniques, samples, and sample evaluations needed to address each information need; and (c) help finance acquisition of the required data and conduct of the analyses.

Analysis Model for Sulfur-Iodine and Hybrid Sulfur Thermochemical Cycles

Abstract In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty due to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. The MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.

SIMULATION OF SULFUR-IODINE THERMOCHEMICAL HYDROGEN PRODUCTION PLANT COUPLED TO HIGH-TEMPERATURE HEAT SOURCE

Abstract This paper presents a transient control volume modeling scheme for both the sulfur-iodine (SI) and Westinghouse hybrid sulfur (HyS) thermochemical cycles. These cycles are very important candidates for the large-scale production of hydrogen in the 21st century. In this study, transient control volume models of the SI and HyS cycles are presented, along with a methodology for coupling these models to codes that describe the transient behavior of a high-temperature nuclear reactor. The transient SI and HyS cycle models presented here are based on a previous model with a significant improvement, namely, pressure variation capability in the chemical reaction chambers. This pressure variation capability is obtained using the ideal gas law, which is differentiated with respect to time. The HyS model is based on a time-dependent application of the Nernst equation. Investigation of the new pressure assumption yields a peak pressure rate of change of 5.877 kPa/s for a temperature-driven transient test matrix and 2.993 kPa/s for a mass flow rate–driven transient test matrix. These high rates of pressure change suggest that an accurate model of the SI and/or HyS cycle must include some method of accounting for pressure variation. The HyS model suggests that the hydrogen production rate is directly proportional to the SO2 production rate.

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

Abstract The sulfur-iodine (SI) cycle is one of the leading candidates in thermochemical processes for hydrogen production. In this paper a simplified model for the SI cycle is developed with chemical kinetics models of the three main SI reactions: the Bunsen reaction, sulfuric acid decomposition, and hydriodic acid decomposition. Each reaction was modeled with a single control volume reaction chamber. The simplified model uses basic heat and mass balance for each of the main three reactions. For sulfuric acid decomposition and hydriodic acid decomposition, reaction heat, latent heat, and sensible heat were considered. Since the Bunsen reaction is exothermic and its overall energy contribution is small, its heat energy is neglected. However, the input and output streams from the Bunsen reaction are accounted for in balancing the total stream mass flow rates from the SI cycle. The heat transfer between the reactor coolant (in this case helium) and the chemical reaction chamber was modeled with transient energy balance equations. The steady-state and transient behavior of the coupled system is studied with the model, and the results of the study are presented. It was determined from the study that the hydriodic acid decomposition step is the rate-limiting step of the entire SI cycle.

MELCOR Simulations of the Severe Accident at the Fukushima Daiichi Unit 1 Reactor

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

Abstract In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code and developing an understanding of the likely accident progression. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of MELCOR and the Fukushima models against plant data. In this study Sandia National Laboratories developed MELCOR 2.1 models of Fukushima Daiichi Units 1 (1F1), 2, and 3 as well as the Unit 4 spent fuel pool. This paper reports on the analysis of the 1F1 accident. Details are presented on the modeled accident progression, hypothesized mode of failures in the reactor pressure vessel (RPV) and containment pressure boundary, and release of fission products to the environment. The MELCOR-predicted RPV and containment pressure trends compare well with available measured pressures. Conditions leading up to the observed explosion of the reactor building are postulated based on this analysis where drywell head flange leakage is thought to have led to accumulation of flammable gases in the refueling bay. The favorable comparison of the results from the analyses with the data from the plant provides additional confidence in MELCOR to reliably predict real-world accident progression. The modeling effort has also provided insights into future data needs for both model development and validation.

Development of design and simulation model and safety study of large-scale hydrogen production using nuclear power.

Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.