Juan J. Carbajo

A review of the thermophysical properties of MOX and UO2 fuels

Abstract A critical review of the thermophysical properties of UO 2 and MOX fuels has been completed, and the best correlations for thermophysical properties have been selected. The properties reviewed are solidus and liquidus temperatures of the uranium/plutonium dioxide system (melting and solidification temperatures), thermal expansion and density, enthalpy and specific heat, enthalpy (or heat) of fusion, and thermal conductivity. Only fuel properties have been reviewed. The selected set of property correlations was compiled to be used in thermal-hydraulic codes to perform safety calculations.

Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

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.

Development of a Twin-Screw D 2 Extruder for the ITER Pellet Injection System

A twin-screw extruder for the ITER pellet injection system is under development at the Oak Ridge National Laboratory. The extruder will provide a stream of solid hydrogen isotopes to a secondary section, where pellets are cut and accelerated with single-stage gas gun into the plasma. A one-fifth ITER scale prototype extruder has been built to produce a continuous solid deuterium extrusion. Deuterium gas is precooled and liquefied before being introduced into the extruder. The precooler consists of a copper vessel containing liquid nitrogen surrounded by a deuterium gas filled copper coil. The liquefier is comprised of a copper cylinder connected to a Cryomech AL330 cryocooler, which is surrounded by a copper coil that the precooled deuterium flows through. The lower extruder barrel is connected to a Cryomech GB-37 cryocooler to solidify the deuterium (at 15 K) before it is forced through the extruder nozzle. A viewport located below the extruder nozzle provides a direct view of the extrusion. A camera is used to document the extrusion quality and duration. A data acquisition system records the extruder temperatures, torque, and speed, upstream, and downstream pressures. This paper will describe the prototype twin-screw extruder and initial extrusion results.

MELCOR sensitivity studies for a low-pressure, short-term station blackout at the Peach Bottom plant

Abstract This paper summarizes the results of analyses performed to assess the effect of a variety of design parameters and operational procedures on a station blackout severe accident at the Peach Bottom Atomic Power Station. The severe-accident melcor code, version 1.8.1 was used in these analyses. The following sensitivity studies were completed: effect of the automatic depressurization system actuation timing on the accident progression; effect of fuel and cladding porosities on vessel failure and containment failure times; effect of several parameters on the amount of in-vessel steel ejected into the cavity after vessel failure; effect of different parameters on vessel penetration failure time; vessel failure timing; and lower plenum shroud and core shroud temperatures. These sensitivity studies provided valuable insights into the melcor code behavior and into the progression of this severe accident. The most significant results are: (a) the optimum steam cooling of the core is accomplished when the automatic depressurization system is actuated when the core water level is at one-third of the active core height, delaying vessel failure by minutes and containment failure by hours, (b) vessel failure is significantly delayed (by 2 h) when lower-plenum debris quenching is included in the model, and (c) the core shroud melts during this transient.

Experimental study of scram transients in generalized liquid-metal fast breeder reactor outlet plenums

An experimental study was made of the transient behavior of the coolant in a generalized outlet plenum after a scram in a liquid-metal fast breeder reactor. Different conditions of fluid flow rate and temperature and plenum geometry were used for a three-exit loop-type reactor system. Water was used to simulate the coolant, and a small-scale model was used to simulate the various generalized plenum configurations. Thermocouples monitored the temperatures at 48 locations within the outlet plenum. The thermohydraulic behavior within the plenum has been studied in detail; analysis of this behavior is described as having three consecutive discrete stages. Correlations found in the literature accurately predict the height of the negatively buoyant plume. New correlations have been obtained.

Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety

This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panels assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

Sodium boiling experiments in the THORS facility

Abstract Experiments have been performed with 19- and 61-pin test assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility at the Oak Ridge National Laboratory (ORNL) since 1971. The THORS Facility is a high-temperature sodium system operated for the US Liquid-Metal Fast Breeder Reactor (LMFBR) Safety Program. The facility is used primarily for testing simulated LMFBR fuel subassemblies (pin (bundles). High-performance, electrically heated fuel pin simulators (FPSs) duplicate the heat generating capabilities and the dimensional characteristics of the nuclear fuel pins. A number of test bundles have been built and operated to obtain base thermal-hydraulic data, inlet and heated zone blockage data, and transient boiling data. Five of these bundles have been operated under two-phase conditions. Sodium boiling for periods up to twelve minutes were sustained in one bundle. (The lengths of the periods were limited only by automatic data recording capability). Clad dryout occurred in several tests. Tests were run at widely varying conditions of flow and power density. Testing with nonuniform power distribution across the bundle was also a part of the program. A 19-pin bundle with 12 peripheral guard heaters and a 6-subchannel blockage around the center pin in the heated zone was tested. The test program for this bundle was designed to determine if local boiling in the wake of the blockage propagates radially or axially during quasi-steady-state conditions. Post-test inspection revealed that significant helical distortion of the FPSs occurred in the vicinity of the blockage plate. This distortion probably influenced the boiling behavior. In the more severe tests, boiling initiated at the outlet of the heated zone and propagated radially into the unblocked subchannels after it had progressed upstream to the blockage. The subchannel analysis codes, SABRE and COBRA, accurately predict the extent of the boiling region. Experimental and analytical studies of sodium boiling behavior in unblocked 19- and 61-pin bundles indicate that cooling can be maintained for a significant period of time beyond boiling inception in a flow-power transient. Quasi-steady-state boiling occurred under natural-convection conditions. Investigations of the temperature data indicate that the thermal-hydraulic behavior during boiling transients is determined by two-dimensional effects, and that one-dimensional models cannot accurately predict the important phenomena associated with sodium boiling in test bundles. The subchannel code SABRE-2P (with a simple two-phase multiplier boiling model) and the two-region equilibrium mixture code THORAX (developed at ORNL) accurately predict the two-dimensional behavior between boiling inception and dryout. Extrapolation of the data from the smaller bundle tests to full-size fuel assemblies shows that the time between boiling inception and dryout would be lower for a 217-pin bundle than for a 61-pin bundle for a comparable transient. However, the time delay would still be significant, especially in a heterogeneous reactor core.

Sodium fast reactor safety and licensing research plan. Volume II.

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Plenum-2A, a program for transient analysis of liquid-metal fast breeder reactor outlet plenums

A new computer program, PLENUM-2A, has been developed to predict the conditions in the outlet plenum of a liquid-metal fast breeder reactor after a transient scram. This program divides the tansient into three main consecutive discrete states. Two additional stages, one occurring before the transiet begins and the other occurring as the transient ends, are also considered. The three main stages are described in detail, and calculations and experimental results are compared. The agreement between the calculations and experimental data is very good.