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DOE OSTI2016

Experimental Study for Two Phase Flow Instability in the PCHE Type Steam Generator in SMR

Conventional U-tube type heat exchangers have some limitations when applied to Small Modular Reactor (SMRs). In particular, the size of steam generators is important in the SMR, as it greatly affects the size, and thus the cost, of the reactor vessel.. Therefore, the circuit heat exchanger (PCHE) is suggested as an alternative. PCHE, a compact type heat exchanger manufactured by HEATRIC, has several advantages such as the reduced size, wide operation range until 900? and 900 bar and high power density due to the large heat transfer area over volume ratio. Recently, it is newly suggested as an application of a steam generator. MIT and KAIST have already conceptually suggested PCHE as a steam generator. Under boiling condition, flow instability is one of the critical issues-resulting in unstable mass flow rate, sudden temperature change and system control failure. According to the flow mechanism, there are various types of instabilities in the heat exchanger and each flow tendency is different. Typical examples are Ledinegg instability of the static instability and pressure-drop oscillation and density-wave oscillation of the dynamic instability. The characteristic of Ledinegg instability is that the flow under unstable condition changes into the stable flow. The characteristic of dynamic instability is that the flow fluctuates continuously. Pressure-drop oscillation has the period of tens of seconds and the system pressure also highly oscillates with the period of flow instability. In comparison, the density-wave oscillation has the period of few seconds and the system pressure is relatively stable. Previous studies on flow instability are focused on the single channel flow and helical tube type heat exchanger. There has been very limited experimental research on flow instability using PCHE. In KAIST, the test facility was constructed to study the thermal hydraulics in PCHE. The objectives of our test facility are: (1) to identify the occurrence of instability under two phase and super-heated steam flow in PCHE, and (2) to examine whether it can operate normally using throttling effect. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Modeling of Two-Phase Flow around Spacers with Mixing Vanes in Reactor Rod-Bundles

Reactor fuel rod bundles experience complex two-phase flow patterns that are governed by multiple interacting phenomena. Proper understand of the underlying flow hydrodynamics can point to potential hotspots within nuclear reactors. The structural supports, called spacer grids, have mixing vanes fixed to the trailing edges to enhance mixing and swirl within the flow. This is an added phenomenon that makes the flows even more complicated to model. As reactors can experience boiling conditions, it is important to understand how the spacer geometry affects local velocity and phase distributions in these geometries. This work represents a continuation of a study that has been focused on developing, testing, and validation of a consistent multidimensional model of dispersed bubbly flow along PWR coolant channels,. The proposed model has been implemented in the NPHASE-CMFD code the results of computer simulations have been compared to experimental results of Yloenen. One of the objectives of the current study has been to examine the effects of selected recent modeling improvements at RPI in NPHASE-CMFD as a part of the continuing model development and validation efforts. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

VUSAT - A VERA-CS Uncertainty and Sensitivity Analysis Toolkit

VERA-CS (the Virtual Environment for Reactor Applications, Core Simulator) is a high-fidelity multiphysics core simulator under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS includes three main components: MPACT for reactor physics and neutron transport, COBRA-TF for thermal-hydraulics, and ORIGEN for isotopic depletion. VERA-CS has been applied to an array of problems including core physics analysis, and the modeling for all of the fuel cycles of Watts Bar Nuclear Unit 1 (WBN1). MPACT is a 3D pin-resolved reactor transport code developed by Oak Ridge National Laboratory (ORNL) and the University of Michigan. COBRA-TF (CTF) is a three-field two-phase flow subchannel analysis code developed by ORNL and North Carolina State University. CTF solves the mass, momentum, and energy equations for the liquid, vapor, and droplet phases for each subchannel in the core and captures the axial flow within each channel as well as cross-flow between channels. Direct coupling between neutronics (MPACT) and thermal-hydraulics (CTF) in VERA-CS provides temperature and density feedback. CTF includes a number of models important for LWR safety and design analysis including flow regime dependent two-phase wall heat transfer, inter-phase heat transfer and drag, droplet breakup, and quench-front tracking. ORIGEN is developed by ORNL and released through the SCALE package. ORIGEN is directly integrated in MPACT through its API to calculate the changes in nuclides for every depletable cross-section region. Uncertainties exist in safety calculations with computer codes for nuclear reactor modeling and simulation, and they come from code limitations, scaling inaccuracies embedded in the experimental data, and uncertainties associated with the state of the reactor at the initiation of the transient. Consequently, for applications of high fidelity calculations using VERA-CS, proper uncertainty quantification (UQ) and sensitivity analysis (SA) of the computing models of reactor neutronics and thermal-hydraulics are important to comprehensively understand the results. New methods in UQ and SA have to be developed in multi-physics calculations using VERA-CS. Toward this end, a software package called VUSAT which stands for VERA-CS Uncertainty and Sensitivity Analysis Toolkit has been developed to facilitate the applications of VERA-CS within the context of best estimate plus uncertainty (BEPU) analysis framework. In this work, the VUSAT application to analyze departure from nucleate boiling (DNB) is presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Uncertainty Quantification and Sensitivity Analysis for Coupled VERA-CS and FRAPCON

The current nuclear power plant fleet in the U.S. is facing numerous challenges economically ranging from low natural gas price to the rapidly growing renewable energy sources such as wind and solar power. In order for the nuclear industry to keep the viability of its operating fleet, new innovations have to be introduced to lower the costs to operate the fleet. One area that could contribute to the cost reduction is to lower the fuel cycle cost and improve the plant safety through initiatives such as higher burnup fuel and accident tolerant fuel. At the same time, the nuclear industry has to address future challenges on the horizon such as load following and other flexible operating strategies as well as those associated with new regulations such as 10 CFR 50.69 and 10 CFR50.46c etc. All these require coupled analysis of core neutronics, thermal hydraulics and fuel performance, which necessitates the development of a multi-physics toolkit. Since different models, assumptions, and fidelities have gone into various computer codes, how to consistently propagate uncertainties in a multi-physics environment warrants careful evaluation. The U.S. DOE's Light Water Reactor Sustainability (LWRS) Program initiated a project to develop analytical capabilities to support the industry to address these challenges. The general idea behind the initiative was the development of an integrated multiphysics evaluation model. This integrated evaluation model is called LOTUS which stands for LOCA Toolkit for the U.S. Light Water Reactors. The focus of LOTUS is to establish the automation of the interface between the five disciplines. LOTUS framework ensures that uncertainties can be easily propagated to solve the problems forward and backward (inverse) with large numbers of simulations. The objective of this work is to perform uncertainty quantification (UQ) and sensitivity analysis (SA) on a coupling of CASL's Virtual Environment for Reactor Applications-Core Simulator (VERA-CS) and NRC's fuel performance code FRAPCON. The work presented herein is a subset of the overall LOTUS framework. The figures of merit (FOM) selected for analysis are minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and the gap conductance at peak power (GCPP). All FOM are obtained at the end of the of a depletion case for a half cycle lasting 263 days. The minimum is taken over all spatial location for each pin at the end of the half cycle. This value is calculated in VERACS alone and has been analyzed in previous work by Brown and Zhang. The MFCT is available in both VERA-CS and FRAPCON, but is calculated via different methods (see Model Inconsistencies section for more details). For this work, only the FRAPCON values are reported due to the code's maturity from extensive validation in fuel performance. Previous work includes SA on a standalone nuclear fuel performance code VTT-modified FRAPCON- 3.4 by Ikonen. The GCPP is the gap conductance at the location corresponding to maximum power in the fuel rod. This value is available in FRAPCON alone and highlights the capability of LOTUS to obtain unique fuel performance metrics for each pin in a core simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Progress in Development of Fluoride-salt-cooled High-Temperature Reactors (FHRs)

Fluoride-salt-cooled High-temperature Reactors (FHRs) are a new class of reactors that use graphite-matrix coated-particle fuel with a clean liquid salt coolant. The largest university FHR program is an Integrated Research Project (IRP) sponsored by the U.S. Department of Energy and others that is led by the Massachusetts Institute of Technology (MIT) and includes the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The year 2016 was the fifth year of this program. We report herein some of the major results of work done in 2016. The base-line FHR concept studied in this IRP uses (1) graphite-matrix coated-particle fuel with a pebble bed geometry and (2) a clean liquid salt coolant-flibe (a mixture of {sup 7}LiF and BeF{sub 2}). The combination of a very high-temperature fuel with failure temperatures above 1650 deg. C, a coolant boiling point above 1400 deg. C, and a coolant that traps most fission products implies the capability to assure no major fuel failure under severe accident conditions. There are a family of FHR concepts using different fuel geometries and different fluoride salts. The power conversion system is a Nuclear Air-Brayton Combined Cycle (NACC) with peak power provided by natural gas or a Firebrick Resistance Heated Energy Storage (FIRES) system. This enables a base-load reactor to provide variable electricity to the grid that substantially increases revenue relative to base-load nuclear reactors. Most of the results reported herein are applicable to most FHR concepts as well as liquid-fueled molten salt reactors (MSRs)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

PLTEMP/ANL Verification for Assemblies Having Five-Layer Fuel Tubes

PLTEMP/ANL v4.2 is a steady-state single-phase (light or heavy water) code used by ANL and by research laboratories in several countries since 1980 in the analysis of various research reactors whose cores contain assemblies made of multiple fuel plates or nested fuel tubes. This analysis is typically needed in the conversion of cores from highly-enriched uranium to low-enriched uranium. PLTEMP calculates temperature distribution and safety margins, i.e., onset of nucleate boiling ratio (ONBR), onset of Ledinegg flow instability ratio (OFIR), and critical heat flux ratio (CHFR). The analytical method in the code for calculating radial and axial temperature distributions in fuel assemblies having multiple nested coaxial fuel tubes was recently extended and verified to allow each tube to have five layers instead of three. The method also works for fuel assemblies having flat fuel plates by setting the radii of curvature of plates to large values ({approx}50 m). The method models heat sources in all five layers and in the coolant channels. To verify the extended method, an Excel spreadsheet was developed using an alternative solution method to calculate the temperature distribution in a reactor of two fuel assemblies, each made of two five-layer flat fuel plates. The code and the spreadsheet solution methods are described and their solutions are compared.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Verification of the Two-Fluid Model in the Spectral Element Code Nek-2P for Adiabatic Flow Conditions

The simulation of multiphase flow is a difficult physical, computational, and numerical problem. Sophisticated tools are necessary to provide results encompassing a wide range of physical phenomena. The development of these tools must be supported by rigorous analysis and ongoing efforts of verification and validation to ensure the accuracy of simulations. The Nek-2P multiphase code is being developed at Argonne with the general goal of simulating a wide range of multiphase flows and the primary goal of simulating boiling flows in LWR fuel assemblies. As part of this, there is an ongoing effort to verify the code and validate the underlying physical models. The objective of this work is to present recent results in the verification of Nek-2P. As a starting point, this work is focused on adiabatic flows for the sake of simplicity with the immediate next planned step to expand the effort to include heat and mass transfer. Details of the verification of the implemented mass fraction transport, momentum conservation, and pressure equations in Nek-2P are presented. Results of simulations are compared to analytic solutions when possible, however due to the complexity of multiphase flows, these are seldom available. To overcome this difficulty, an independent model is used to provide solutions for comparison.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Analysis of COBRA-TF Critical Heat Flux Models for Triangular Fuel Assembly Pitch

The Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF) code is a best estimate code which is used to perform thermal-hydraulic analyses of a light water reactor vessel. The COBRA-TF heat transfer package has models calculating the critical heat flux (CHF). In order to verify code's predictive capability for CHF for triangular nuclear fuel assembly pitch, experiment results and system parameters of KS-1 test facility that is single loop model of the VVER primary system located in Kurchatov Institute were simulated using COBRA-TF. In order to calculate CHF for triangular nuclear fuel assembly pitch there is a look-up table generated. In addition, there are some correction factors for critical heat flux values listed the table. In this study, once KS-1 experiment results were simulated using COBRA-TF, correction factors were used to correct CHF results from COBRA-TF calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

CFD Simulation of Bubble Growing Under a Downward Facing Surface

The heat removing of downward facing surface is an important issue in many areas, including nuclear safety, solar storage system and electronic cooling devices. Therefore, the characteristic and heat removal capability of downward facing boiling were investigated by many experiments in recent years. Among these studies, the thickness of bubble is considered as an important parameter. The increasing of bubble thickness will increase the thermal resistance of cooling process and reduce the cooling ability. In order to focus on the dynamics of bubble thickness, Das et al. injected air into sodium dodecyl sulphate (SDS) solution through a nozzle under the horizontal downward facing surface. The bubble will be formed by the air injection and kept under the horizontal plate. The diameter of nozzle, horizontal plate and SDS solution tank are 2 mm, 20 mm and 30 mm, respectively. The growing processing and the stable thickness of the bubble were recorded in the experiment. Basing on the conditions of Das's experiment, the appropriate computational fluid dynamics (CFD) simulation was conducted. The growing process and the stable thickness of bubble under difference contact angle are predicted by CFD in this study.

42 ENGINEERING↗

DOE OSTI2017

Simulating Bubble Deformations Using the Intersection Marker (ISM) Front-Tracking method

Surface and volume tracking methods have been under continuous development over the last 50+ years since the first MAC (Marker and Cell) and SLIC (Simple Line Interface Calculation) methods appeared in the 1960's. The sustained interest in simulating multiphase flow stem from their many practical applications in the physical world; from biological systems, to chemical systems, to boiling dynamics and atmospheric heat and mass transfer. The ability to capture the details of interface evolution as well as the physics of interfacial heat and mass transfer has thus been a tempting goal for many numerical researchers. Numerical analysts, apart from the well-established and popular VOF PLIC (Volume of Fluid, Piecewise Linear Interface Construction) method, can now call upon even more computationally intensive options such as the LS (Level Set) method, the FT (Front Tracking) method, the MoF (Moment of Fluid) method and hybrid methods.

42 ENGINEERING↗

DOE OSTI2017

Validation of CTF Droplet Entrainment and Annular/Mist Closure Models using Riso Steam/Water Experiments

This document summarizes the work to validate the droplet entrainment and de-entrainment models, as well as two-phase closure models in the CTF (Coolant Boiling in Rod Arrays - Two Fluids) code by comparison with experimental data obtained at Risoe National Laboratory. CTF is the modernized version of the subchannel code, COBRA-TF, being jointly developed by Oak Ridge National Laboratory and North Carolina State University for applications in the Consortium for Advanced Simulation of Light Water Reactors (CASL). The Risoe data included a series of more than 250 steam/water experiments that were performed in both tube and annulus geometries over a range of various pressures and outlet qualities. Experimental conditions were set so that the majority of cases was in the annular/mist flow regime. Measurements included liquid film flow rate, droplet flow rate, film thickness, and two-phase pressure drop. CTF was used to model 180 of the tubular geometry cases, matching experimental geometry, outlet pressure, and outlet flow quality to experimental values. CTF results were compared to the experimental data at the outlet of the test section in terms of vapor flow fractions, and entrained liquid flow fractions, and pressure drop per unit length.

42 ENGINEERING↗

DOE OSTI2017

Review of CTF's Fuel Rod Modeling Using FRAPCON-4.0's Centerline Temperature Predictions

Coolant Boiling in Rod Arrays-Two Fluid (COBRA-TF), or CTF1, is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF's fuel rod modeling is originally developed for VIPRE code. Its methodology is based on GAPCON and FRAP fuel performance codes, and material properties are included from MATPRO handbook. This work focuses on review of CTF's fuel rod modeling to address shortcomings in CTF's temperature predictions. CTF is compared to FRAPCON which is U.S. NRC's steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes in fuel rod variables and temperatures including the effects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite difference or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF's dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF's models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF's dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF's dynamic gap conductance model. Improving the CTF's dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Single Rod Reflood Quenching Experiments with Simulated Decay Heat for Accident-Tolerant Fuel Claddings

The Loss-of-Coolant Accident (LOCA) is a hypothetical situation which includes a sudden loss of coolant, overheating of fuel rods, and emergency coolant injection. This situation could be modeled as a quenching process of cladding material into the coolants, a reverse process of boiling the coolants. The KAERI have been developing a CrAl-coating type Accident-Tolerant Fuel (ATF) claddings to enhance safety margin in such severe accident situations and evaluations for the performance in various aspects are on process. We established a single-rod type reflood quenching experiment setup to simulate the LOCA process. Unlike other conventional quenching experiment setup, our experiment setup is capable of heating during the reflooding process, simulating the decay heat in the real core. We installed heaters and thermocouples (TC) inside the cladding specimens to minimize any disturbances on the interfacial layer during the experiment. Currently the setup is on test run and we present experiment data and verification with conventional Zircaloy-4 claddings. We are going to conduct the reflood quenching experiment with KAERI's CrAl-coated ATF claddings to evaluate the cooling performances.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Development of a Multiphase Adjoint Capability in OpenFOAM

Thermal hydraulic predictions of flow regimes within nuclear reactor cores require significant computational resources and accuracy to ensure that the core design does not violate the thermal limits of materials. Multiphase computational fluid dynamics (CFD) has become a valuable computational capability for predicting flow characteristics as they relate to applications such as nuclear reactor safety and mixing phenomena around spacer grids within fuel assemblies. Multiphase CFD requires closure relationships for both the effects of turbulence as well as the transfer of mass, momentum, and energy between phases. Understanding the sensitivity of the parameters used in closure relationships is important for improving accuracy, but performing rigorous sensitivity analysis is computationally expensive. Adjoint mathematics have been utilized for reactor physics and fluid dynamics applications with respect to verification, sensitivity analysis, adaptive mesh refinement, and optimization. The usage of adjoint methods to predict the sensitivities of a variety of closure relationship parameters within a given problem has been shown to return sensitivity responses with significantly fewer computations than normal 'large-scale' methods. As the complexity of multiphase CFD methods increases, the demand for accurate verification tools and improvements in computational efficiency also increases, making the development of adjoint capabilities an attractive option. Closure relationships that describe interphase momentum physics in multiphase CFD modelling rely heavily on prescribed constants determined from experimental data. Due to the computational efficiency of an adjoint approach, the time to derive these closure relationships could be dramatically reduced, and data assimilation techniques could be used to return probability density functions of closure models derived using experimental data. This work develops a multiphase adjoint capability for the Eulerian-Eulerian two field, two phase equations. Specifically, the adjoint capability has been developed for the sub-cooled nucleate boiling solver boilEulerFoam developed by Dr. Nam Dinh of NC State University within the open-source CFD library OpenFOAM.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Toward High Fidelity LES Simulations of 5 x 5 PWR Fuel Bundle with Mixing Vane

The present work illustrates one example of a computational fluid dynamics (CFD) application in the nuclear engineering field. Since CFD is one of the most powerful tools available at the actual state of the art that could in principle help and assist the design process, a compromise should be established between the accuracy of the CFD numerical model (RANS - LES - DNS) and the computational cost of the simulations. In this spirit, this work will present the computational results from two different CFD model categories, Large Eddy Simulations (LES) and Reynolds-averaged Navier-Stokes (RANS) models. Both treat the flow turbulence issue in two separate ways, which differ mostly in the resolution of the turbulence statistics and the computational cost. The single-phase hydraulic problem presented involves the steady-state turbulent flow field modelling and resolution around a scaled PWR mixing vane grid, designed by Westinghouse and tested by Texas A and M University. The role of the mixing vane/spacer grid is to enhance the turbulent mixing and hence heat transfer of the core coolant around fuel bundles, thus increasing the departure of nucleate boiling-critical heat flux (DNB-CHF) point, as well as important structural functions for the rods bundles. The authors numerically studied the flow field characteristics at the region downstream of the spacer grids and validated their results to particle imagery velocimetry (PIV) measurements. Dominguez and Hassan performed two-dimensional time resolved PIV (TR-PIV) measurements within a 5 x 5 PWR rod bundle with spacer-grids that utilized a matching index of refraction (MIR) approach. Their averaged data showed a rather uniform direction velocity profile after the spacer grid while the downstream illustrated a no slip wall profile. The available PIV flow field experimental results from Texas A and M show complex structures of the flow field downstream the mixing vanes with the formation of secondary flows that tend to enhance the turbulent heat transfer, due to the mixing vanes geometry itself. Computational results will be validated with the available experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Coupled CFD-DEM Analysis of Molten Salt-Cooled Pebble-Bed Reactor Experiment

The development of new nuclear reactor designs is an area of ongoing investigation, and a focus of dedicated work for various laboratories around the world. Despite the steadfast and reliable service of Light Water Reactors (LWR), the ideal power reactor, with a combination of high operational temperature, passive safety, proliferation resistance, economic build and waste minimization remains elusive. One promising design that might address these diverse, and at times contradictory requirements, is the Molten Salt Reactor (MSR). In its most radical form, a liquid fueled (LF) MSR has the nuclear fuel dissolved in the molten-salt primary coolant itself, as demonstrated by Oak Ridge National Lab (ORNL) in 1965 to 1969. An alternative fuel arrangement is to retain conventional, solid fuel elements. A proposed method is to implement pebble fuel elements, as have been used in previous gas-cooled reactors, with the gas coolant being replaced with a molten salt. This is referred to as pebble-bed, molten-salt reactor (PB-MSR). The spherical fuel pebbles used within PB-MSRs contain TRISO particles dispersed throughout the pebble's graphite matrix. Many pebbles are constrained to form a porous bed within the core region of a reactor. Coolant molten-salt passes through the pebble bed to remove heat produced due to the nuclear fission taking place within the fuel pebbles' TRISO particles. Various benefits are apparent with the PB-MSR design. When considering the choice of fuel elements, they are a geometry that has been produced and utilized in a similar way in the past, within gas-cooled reactors. The fuel pebbles are also able to be added to, and removed from, the core during operation, enabling on-line refueling. When considering the coolant choice, molten-salts have much higher boiling points (1703 K at 1 atm. for FLiBe) than conventional coolants such as water (630 K at 150 atm.), and are capable of reaching these temperatures at atmospheric pressure. This in turn removes the need for high pressure containment that is present in light water reactors and gas-cooled reactors. The future implementation of new reactors, and the search for improvements in efficiency and safety in current designs, require the advancement of simulation tools, which this work ultimately aims to address.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Evaluation of an unexpected reflector temperature effect for water-isolated array configurations - 14500

It has been conventional wisdom in the criticality safety discipline that for spent fuel systems like water pit storage and transportation, modeling water at its peak density will maximize reactivity. Thus, modeling water at 3.96 deg. C will produce a conservatively bounding k{sub eff} for that configuration. This assumption has been found to be incorrect for certain high-leakage water-isolated array configurations. In fact, temperatures up to the boiling point of water can in fact be more reactive, and the most reactive temperature is highly dependent upon the specific fuel configuration being analyzed. In light of the accident at Fukushima-Daiichi - where the spent fuel pool temperature increased - knowledge of this unexpected behavior is essential to ensuring that criticality safety evaluations are conservatively bounding. To illustrate the reflector temperature effect, MC21 calculations are provided for a sensitivity study based on LEU-COMP-THERM-044, which models the IPEN/MB-01 reactor at various temperatures between 3.9 deg. C and 93.3 deg. C. (authors)

42 ENGINEERING↗

DOE OSTI2015

Geyser eruption intervals and interactions: Examples from El Tatio, Atacama, Chile

Here, we compare and contrast data collected in 2012 and 2014 from the El Tatio geyser field, Chile. We identify geyser systems that evolve over time, including changes in the interval between eruptions, development of new thermal features, and interactions between geysers. We study three different cases: (a) an isolated geyser, which is periodic and has nearly identical eruptions every cycle; (b) a geyser and coupled noneruptive pool, where the geyser has nonregular cycles and several preplay eruptions before the main eruption; and (c) two geysers and a mud volcano, which have nonregular cycles and are all interacting. Though geysers erupt with different styles, we recognize some common features: the conduit recharges with liquid during the quiescent period, bubbles enter the conduit before eruptions, and eruptions occur when water boils in the upper part of the conduit. The episodic addition of heat may govern the periodicity, while the depth where heat is added dictates the eruption style: conduits with deeper heat input are more likely to show preplay or minor eruptions. The interactions between thermal features can be explained by pressure transmission in subsurface permeable layers between geyser conduits.

58 GEOSCIENCES↗