Larry J. Ott

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.

STUDY OF FUKUSHIMA DAIICHI NUCLEAR POWER STATION UNIT 4 SPENT-FUEL POOL

A study on the Fukushima Daiichi nuclear power station spent-fuel pool (SFP) at Unit 4 (SFP4) is presented in this paper. We discuss the design characteristics of SFP4 and its decay heat load in detail and provide a model that we developed to estimate the SFP evaporation rate based on the SFP temperature. The SFP level of SFP4 following the March 11, 2011, accident is predicted based on the fundamental conservation laws of mass and energy. Our predicted SFP level and temperatures are in good agreement with measured data and are consistent with Tokyo Electric Power Company evaluation results.

Advanced severe accident response models for BWR application

Abstract Boiling Water Reactors (BWRs) and their primary containments have unique features that must be modeled for best estimate severe accident analysis. The BWRSAT Program at ORNL has developed and incorporated into its calculational code suite several advanced response models for application in BWR severe accident analyses. Major features of these models include representation of (1) heat transfer to all in-core structures, (2) the effect of SRV actuations, (3) structural/steam chemistry effects (4) progressive core structural relocation (including “candling”), (5) failure of coreplate and formation of debris bed in reactor bottom head, (6) vessel penetration faiure, and (7) the time dependent egress of core debris from the vessel. This paper discusses these advanced models and presents the results of calculations utilizing the models for the Station Blackout Severe Accident Sequence at Browns Ferry.

Overview of materials technologies for space nuclear power and propulsion

A wide range of different space nuclear systems are currently being evaluated as part of the DOE Special Purpose Fission Technology program. The near-term subset of systems scheduled to be evaluated range from 50 kWe gas-, pumped liquid metal-, or liquid metal heat pipe-cooled reactors for space propulsion to 3 kWe heat pipe or pumped liquid metal systems for Mars surface power applications. The current status of the materials technologies required for the successful development of near-term space nuclear power and propulsion systems is reviewed. Materials examined in this overview include fuels (UN, UO2, UZrH), cladding and structural materials (stainless steel, superalloys, refractory alloys), neutron reflector materials (Be, BeO), and neutron shield materials (B4C,LiH). The materials technologies issues are considerably less demanding for the 3 kWe reactor systems due to lower operating temperatures, lower fuel burnup, and lower radiation damage levels. A few reactor subcomponents in the 3 kWe reactors...

BWRSAR calculations of reactor vessel debris pours for peach bottom short-term station blackout

Abstract This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident Response (BWRSAR) code and are based upon consideration of the Peach Bottom Atomic Power Station. The modeling strategies employed within BWRSAR for debris relocation within the reactor vessel are briefly discussed and the calculated events of the accident sequence, including details of the calculated debris pours, are presented.

Small scale BWR core debris eutectics formation and melting experiment

Abstract A small scale experiment has recently been performed at Oak Ridge under the auspices of the Boiling Water Reactor Severe Accident Technology (BWRSAT) program to provide information concerning the formation of mixtures during heatup of representative BWR reactor vessel bottom head debris and to determine the composition and melting temperatures of these mixtures. The initial structure of the bottom head debris layers modeled in the experiment was taken from the results of recent Boiling Water Reactor Severe Accident Response (BWRSAR) code predictions for the short-term station blackout accident sequence. The experimental results provide useful information concerning the mixtures formed and their proportions and properties. The observed run-off of a stainless steel-zirconium eutectic alloy supports the contention that the initial pour from a BWR reactor vessel would consists of molten metals at relatively low temperatures.

Insight from Fukushima Daiichi Unit 3 Investigations using MELCOR

Abstract During the emergency response period of the accidents that took place at the Fukushima Daiichi nuclear power plant (NPP) in March of 2011, researchers at Oak Ridge National Laboratory (ORNL) conducted a number of studies using the MELCOR code to help understand what was occurring and what had occurred. During the postaccident period, the U.S. Department of Energy (DOE) and the U.S. Nuclear Regulatory Commission (NRC) jointly sponsored a study of the Fukushima Daiichi NPP accident with collaboration among ORNL, Sandia National Laboratories, and Idaho National Laboratory. The purpose of the study was to compile relevant data, reconstruct the accident progression using computer codes, assess the codes’ predictive capabilities, and identify future data needs. The current paper summarizes some of the early MELCOR simulations and analyses conducted at ORNL of the Fukushima Daiichi NPP Unit 3 (1F3) accident. Extended analysis and discussion of the 1F3 accident are also presented taking into account new knowledge and modeling refinements made since the joint DOE-NRC study.

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

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.

Sodium Fast Reactor Fuels and Materials: Research Needs.

An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

Overview of Fuel Rod Simulator Usage at ORNL

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out‐of‐reactor experimental facilities to resolve thermal‐hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss‐of‐coolant accident (LOCA) to basic heat transfer research in gas‐ or sodium‐cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized‐water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.