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DOE OSTI2015

Light Water Reactor Sustainability Accomplishments Report

Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you have any questions about the information in the report, or about the LWRS Program, please contact me, Richard A. Reister (the Federal Program Manager), or the respective research pathway leader (noted on pages 26 and 27), or visit the LWRS Program website (www.inl.gov/lwrs). The annually updated Integrated Program Plan and Pathway Technical Program Plans are also available for those seeking more detailed technical Information.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A User Guide to PARET/ANL (V.7.6) (Rev.1)

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA, and has been used by the Reactor Conversion Program to determine the expected transient behavior of a large number of reactors. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Imaging the Core of Fukushima Reactor with Muons

The 9.0-magnitude earthquake, followed by the vast tsunami on March 11, 2011, caused a nuclear crisis at the Fukushima Daiichi reactors. Damage of the reactor cores attracted worldwide attention to the issue of the fundamental safety of nuclear energy. The Japanese government announced a cold shutdown in December 2011, and began a new phase of cleanup and decommissioning. However, it is difficult to plan the dismantling of the reactors without a realistic estimate of the extent of the damage to the cores and knowledge of the location of the melted fuel. Access to the reactor buildings is very limited due to high radiation fields. Los Alamos researchers have used experiments and modeling to show how muon imaging with detectors external to the reactors can enable damage assessment inside Fukushima. Physical Review Letters, AIP Advances, and the Journal of Applied Physics have published different aspects of this work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Embedded Volttron Specification - Benchmarking Small Footprint Compute Devices for Volttron

An embedded system is a small footprint computing unit that typically serves a specific purpose closely associated with measurements and control of hardware devices. These units are designed for reasonable durability and operations in a wide range of operating conditions. Some embedded systems support real-time operations and can demonstrate high levels of reliability. Many have failsafe mechanisms built to handle graceful shutdown of the device in exception conditions. The available memory, processing power, and network connectivity of these devices are limited due to the nature of their specific-purpose design and intended application. Industry practice is to carefully design the software for the available hardware capability to suit desired deployment needs. Volttron is an open source agent development and deployment platform designed to enable researchers to interact with devices and appliances without having to write drivers themselves. Hosting Volttron on small footprint embeddable devices enables its demonstration for embedded use. This report details the steps required and the experience in setting up and running Volttron applications on three small footprint devices: the Intel Next Unit of Computing (NUC), the Raspberry Pi 2, and the BeagleBone Black. In addition, the report also details preliminary investigation of the execution performance of Volttron on these devices.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the power coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.

Connaway, H. M.↗
DOE OSTI2015

Mitigation of Syngas Cooler Plugging and Fouling

This Final Report summarizes research performed to develop a technology to mitigate the plugging and fouling that occurs in the syngas cooler used in many Integrated Gasification Combined Cycle (IGCC) plants. The syngas cooler is a firetube heat exchanger located downstream of the gasifier. It offers high thermal efficiency, but its’ reliability has generally been lower than other process equipment in the gasification island. The buildup of ash deposits that form on the fireside surfaces in the syngas cooler (i.e., fouling) lead to reduced equipment life and increased maintenance costs. Our approach to address this problem is that fouling of the syngas cooler cannot be eliminated, but it can be better managed. The research program was funded by DOE using two budget periods: Budget Period 1 (BP1) and Budget Period 2 (BP2). The project used a combination of laboratory scale experiments, analysis of syngas cooler deposits, modeling and guidance from industry to develop a better understanding of fouling mechanisms and to develop and evaluate strategies to mitigate syngas cooler fouling and thereby improve syngas cooler performance. The work effort in BP 1 and BP 2 focused on developing a better understanding of the mechanisms that lead to syngas cooler plugging and fouling and investigating promising concepts to mitigate syngas cooler plugging and fouling. The work effort focused on the following: • analysis of syngas cooler deposits and fuels provided by an IGCC plant collaborating with this project; • performing Jet cleaning tests in the University of Utah Laminar Entrained Flow Reactor to determine the bond strength between an ash deposit to a metal plate, as well as implementing planned equipment modifications to the University of Utah Laminar Entrained Flow Reactor and the one ton per day, pressurized Pilot Scale Gasifier; • performing Computational Fluid Dynamic modeling of industrially relevant syngas cooler configurations to develop a better understanding of deposit formation mechanisms; • performing Techno-Economic-Analysis for a representative IGCC plant to investigate the impact on plant economics, in particular the impacts on the Cost of Electricity (COE), due to plant shutdowns caused by syngas cooler plugging and fouling and potential benefits to plant economics of developing strategies to mitigate syngas cooler fouling; and • performing modeling and pilot scale tests to investigate the potential benefits of using a sorbent (fuel additive) to capture the vaporized metals that result in syngas cooler fouling. All project milestones for BP 1 and BP 2 were achieved. DOE was provided a briefing on our accomplishments in BP1 and BP2 and our proposed plans for Budget Period 3 (BP 3). Based on our research the mitigation technology selected to investigate in BP 3 was the use of a sorbent that can be injected into the gasifier with the fuel slurry to capture vaporized metals that lead to the deposit formation in the syngas cooler. The work effort proposed for BP 3 would have focused on addressing concerns raised by gasification industry personnel for the impacts on gasifier performance of sorbent injection, so that at the end of BP 3 the use of sorbent injection would be at “pre-commercial” stage and ready for use in a Field Demonstration that could be funded by industry or DOE. A Budget Continuation Application (BCA) was submitted to obtain funding for BP3 DOE but DOE chose to not fund the proposed BP3 effort.

47 OTHER INSTRUMENTATION↗
DOE OSTI2015

Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The Bumpy Torus Experiment

This document summarizes the Bumpy Torus Experiment as a viable fusion reactor concept. Conclusions reached include the following: In 30 years, order-of-magnitude technological advances have occurred in multiple areas of plasma heating and confinement. The ORNL bumpy torus of the 1970s was technology limited. Now that ITER is technology limited, an alternate concept is needed. A device built on such a concept should be current free, CW, modular, have a gentle shutdown, and demonstrable stability. The bumpy torus meets or has the potential to meet all of these criteria. Earlier, stability was not possible due to power limits; it has not been fully tested. It is time to revisit the bumpy-torus concept with a modest new machine.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

System Study: Residual Heat Removal 1998–2013

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and post-trip shutdown-cooling) at 104 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RHR results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Analysis of Loss-of-Offsite-Power Events 1998–2013

Loss of offsite power (LOOP) can have a major negative impact on a power plant’s ability to achieve and maintain safe shutdown conditions. Risk analyses suggest that loss of all alternating current power contributes over 70% of the overall risk at some U.S. nuclear plants. LOOP event and subsequent restoration of offsite power are important inputs to plant probabilistic risk assessments. This report presents a statistical and engineering analysis of LOOP frequencies and durations at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience during calendar years 1997 through 2013. Frequencies and durations were determined for four event categories: plant-centered, switchyard-centered, grid-related, and weather-related. The emergency diesel generator failure modes considered are failure to start, failure to load and run, and failure to run more than 1 hour. The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. No statistically significant trends in LOOP frequencies over the 1997–2013 period are identified. There is a possibility that a significant trend in grid-related LOOP frequency exists that is not easily detected by a simple analysis. Statistically significant increases in recovery times after grid- and switchyard-related LOOPs are identified.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2012

2012 Joint Research Target (JRT) Final Report

The report summarizes: [1] Diagnostic upgrades and analysis improvements that support the JRT; [2] Descriptions of new experiments and a brief summary of findings; [3] New analysis of previous experiments; [4] Plans for further data analysis and recommendations for further work; [5] Scientific publications that are derived from or contributed directly to the JRT. The report is organized as follows: Section 1: Describes inter-machine comparisons that were facilitated by the JRT focus. Similarities and differences found on the different facilities are outlined, and plans for additional analysis of data and supporting simulations are described.Section 2: Reviews diagnostic development, experiments and results from C-Mod. C-Mod dedicated 13.1 run days to the JRT divided among 10 experimental proposals. These were organized into three general thrust areas that represent distinct experimental approaches to realizing the regimes called for in the JRT description. Initial comparisons with linear and nonlinear simulation have been carried out for these experiments, and plans for an extensive campaign of analysis have been outlined. Section 3: Reviews results from DIII-D including new experiments, detailed comparisons of simulations to previously collected data and connections between the new work and past research. Four experimental days were dedicated to the JRT in 2012. These included studies of L-modes, H-modes and QH-modes. Section 4: Summarizes new analysis of data collected before the NSTX shutdown. The work focuses on the roles of low and high-k turbulence; collisionality scans and impurity particle transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

“Exploring High Energy Interactions with CMS at the LHC”

This High Energy Physics research project achieved its goal of exploring high-energy interactions with 7, 8 and 13 TeV data accumulated by CMS at the Energy Frontier. For the original hadron calorimeter (HCAL) and for its upgrade during Long Shutdown 1 (LS1), the PI helped propose and implement the upgrading the phototubes, new electronics, and fast timing of the hadronic forward (HF) and hadronic outer (HO) calorimeters of CMS, projects which he had forcefully advocated since the inception of CMS. The PI and his colleagues Prof. J. Rohlf and chief electronics engineer E. Hazen, his post-docs A. Heister and S. Girgis, and his graduate students (P. Lawson and D. Arcaro) contributed software tools used in perfecting of μTCA and Advanced Mezzanine Card (AMC13) electronics, the PC board that provides clock, timing and DAQ service for HCAL (and now many other subdetectors and central systems in the upgraded CMS detector). This Task reaped the benefits of these hardware contributions 1) to hermiticity for missing energy searches, and 2) to forward tagging jets for Vector Boson Fusion processes by analyzing and publishing early data, including that for the Higgs discovery and for exotic and supersymmetric searches.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

Run 16, eIPM Summary

Three problems with the eIPMs were corrected during the 2015 summer shutdown. These involved ac coupling and 'negative profiles', detector 'dead zone' created by biasing, and gain control on ramp. With respect to Run 16, problems dealt with included gain depletion on horizontal MCP and rf pickup on profile signals; it was found that the MCP was severely damaged over part of the aperture. Various corrective measures were applied. Some results of these measured obtained during Run 16 are shown. At the end of Run 16 there was a three-­day beam run to study polarized proton beams in the AGS. Attempts to minimize beam injection errors which increase emittance by using the eIPMs to measure the contribution of injection mismatch to the AGS output beam emittance are recounted. .

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Specification of Fenix MPI Fault Tolerance library version 1.0.1

This document provides a specification of Fenix, a software library compatible with the Message Passing Interface (MPI) to support fault recovery without application shutdown. The library consists of two modules. The first, termed process recovery , restores an application to a consistent state after it has suffered a loss of one or more MPI processes (ranks). The second specifies functions the user can invoke to store application data in Fenix managed redundant storage, and to retrieve it from that storage after process recovery.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2016

Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping and pitting degradation on the YA-M fuel elements. In the case of scalloping (horseshoeing) a surprising similarity of that defect to those appearing on aluminum plate rolled in over-lubrication conditions, were established. In turn, this made us think that the principal feature responsible for the appearance of these defects, was horizontal cuts in the beryllium reflector block created to arrest the propagation of large vertical crack(s) in Be in PALM cycles with higher overall fluence. This assumption was fully confirmed by the results of thermo-hydraulic simulations. The neutronics data for these modeling experiments were provided using advanced irradiation simulations (MCNP, HELIOS). In the case of pitting erosion the following corrective measures were proposed based upon the results of JMatPro v.8.2 modeling (TTT- and CCT-diagrams): change the fabrication process by adding blister anneal before program anneal, immediately after cold rolling of AA6061plate. This step will allow achieving complete recrystallization, eliminating of strengthening due to metastable precipitates, and reduce the possibility of forming sharp microstructural features upon the surface.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

W Boson Mass Measurement at CDF

This is the closeout report for the grant for experimental research at the energy frontier in high energy physics. The report describes the precise measurement of the W boson mass at the CDF experiment at Fermilab, with an uncertainty of ≈ 12 MeV, using the full dataset of ≈ 9 fb -1 collected by the experiment up to the shutdown of the Tevatron in 2011. In this analysis, the statistical and most of the experimental systematic uncertainties have been reduced by a factor of two compared to the previous measurement with 2.2 fb -1 of CDF data. This research has been the culmination of the PI's track record of producing world-leading measurements of the W boson mass from the Tevatron. The PI performed the first and only measurement to date of the W boson mass using high-rapidity leptons using the D0 endcap calorimeters in Run 1. He has led this measurement in Run 2 at CDF, publishing two world-leading measurements in 2007 and 2012 with total uncertainties of 48 MeV and 19 MeV respectively. The analysis of the final dataset is currently under internal review in CDF. Upon approval of the internal review, the result will be available for public release.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2015

Proppant backflow: Mechanical and flow considerations

One of the concerns of using proppant in geothermal wells, and particularly in enhanced geothermal systems, is proppant flowback. Particulate proppant maintain post-closure conductivity in hydraulically opened fractures. If that proppant is displaced from the near-wellbore region, either due to overflushing during stimulation or flowback to the wellbore at any time, the reduced fracture width chokes the injection or production. Two intermediate-scale laboratory analogs of a propped hydraulic fracture were prepared, and fluid was flowed through a normally stressed, propped fracture into a central wellbore. The tests were conducted in a polyaxial load frame. Acoustic/microseismic activity was measured during the injection programs. In one scenario—radial flow through a transverse fracture to a wellbore—the results suggest the creation of flow channels and nominally intact propped zones around the channels, maintaining fracture aperture. In the other—linear flow through a longitudinal fracture into a wellbore—there was substantially more proppant removal. The measurements have shown a greater tendency for proppant flowback in a linear flow situation (proppant movement is kinematically more restricted for radial convergent flow). The pressure gradients causing flow are exceedingly small and restraining flowback will be difficult. Convergent flow relationships could be an issue for injector wells, which will experience fluid flowback during hard shutdowns.

15 GEOTHERMAL ENERGY↗