Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Design Basis”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 253 records · Page 14

Validation of a plant dynamics code for 4S - Test analysis of natural circulation behavior

A plant transient dynamics code for a sodium-cooled fast reactor was developed by Toshiba. The code is used to evaluate the safety performance of Super-Safe, Small, and Simple reactor (4S) for Anticipated Operational Occurrences (AOOs), Design Basis Accident (DBA) and Beyond DBA (BDBA). The code is currently undergoing verification and validation (V and V). As one of the validation, test analysis of the Shutdown Heat Removal Test (SHRT)-17 performed in the Experimental Breeder Reactor (EBR)-II was conducted. The SHRT-17 is protected loss of flow test. The purpose of this validation is to confirm capability of the code to simulate natural circulation behavior of the plant. As a result, good agreements are shown between the analytical results and the measured data which were available from instrumented subassembly. The detailed validation result of the natural circulation behavior is described in this paper. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Risk assessment of safety-critical data communication in digital safety feature control system - 205

As one of the safety-critical systems in Advanced Power Reactor-1400 (APR-1400) nuclear power plant (NPP), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for transmitting safety component actuation to effectively facilitate various safety-critical field controllers. However, data communication failure risk in the ESF-CCS has yet to be fully quantified, the probability that a safety-critical system in digitalized NPP becomes unsafe due to a network failure must be evaluated to quantify the risk of digital I and C system. In this study, a fault tree model was developed to assess the data link and data network failure-induced unavailability of a safety-critical system function used to generate an automated control signal for actuating design basis accident mitigation equipment. Especially, the risk of data communication failure in a digital safety feature control system was analyzed in consideration of interconnection between safety-critical controllers and the fault-tolerant algorithm implemented in the network configuration. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to safety data link and network failure and their risk effect on safety signal generation unavailability. This study is expected to provide risk information on the safety-critical data communication in a digitalized NPP instrumentation and control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Condensation model for the ESBWR passive condensers

In the General Electric's Economic simplified boiling water reactor (GE-ESBWR) the passive containment cooling system (PCCS) plays a major role in containment pressure control in case of an loss of coolant accident. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure following a design basis accident. There are three PCCS condensation modes depending on the containment pressurization due to coolant discharge; complete condensation, cyclic venting and flow through mode. The present work reviews the models and presents model predictive capability along with comparison with existing data from separate effects test. The condensation models in thermal hydraulics code RELAP5 are also assessed to examine its application to various flow modes of condensation. The default model in the code predicts complete condensation well, and basically is Nusselt solution. The UCB model predicts through flow well. None of condensation model in RELAP5 predict complete condensation, cyclic venting, and through flow condensation consistently. New condensation correlations are given that accurately predict all three modes of PCCS condensation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Counterpart Test on IBLOCA of 13% Cold Leg Break in ATLAS Facility

In the analysis for Loss-of-Coolant Accident (LOCA) as a design-basis accident (DBA) of a pressurized water reactor (PWR), redefinition of the break size based on the risk-informed regulation (RIR) is being extensively investigated due to the potential safety benefits. The pipe integrity and break frequency has been found to depend on the pipe size, so that a complete rupture of a smaller pipe such as a pressurizer surge-line or an emergency core cooling system (ECCS) nozzle is more probable than a double-ended guillotine break (DEGB) of primary coolant circuit piping. The scenario with the break of these pipes can be regarded as intermediate break (IB) LOCA. With an aim of expanding code validation to address safety issues of the IBLOCA and developing an evaluation model (EM) for full spectrum LOCA, the database for the IBLOCA test in an integral effect test (IET) facility is essential. In particular, an experimental data from the counterpart test can contribute to enhance our understanding on the phenomena and resolve the scaling issues in the scaled-down test facilities. The objective of this study is to investigate thermal hydraulic phenomena during the IBLOCA by the counterpart test for the 13% cold leg break in ATLAS facility, which was the A.2 test in the OECDATLAS project. This integral effect test would contribute to evaluate the scaling effect of the ATLAS design and test condition.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Westinghouse Accident Tolerant Fuel Program

The Westinghouse accident tolerant fuel (ATF) program is aimed at providing a significant increase in fuel robustness able to increase its tolerance to beyond design basis accidents at an economically attractive price to nuclear power providers. The Westinghouse ATF program, which began in 2004, advanced significantly during the current first phase (1A and 1B) begun in 2012 of this Department of Energy (DOE) program. Phase 1B will culminate in the manufacture of fuel rods that will undergo a 6 year exposure at steady state conditions in the Advanced Test Reactor and the Halden test reactor (Phase 2) and in transient tests in TREAT and Halden. The goal is to load lead test rods in a commercial reactor in 2022. In this paper, the current status of the ATF development and future irradiation plan is discussed in detail. Westinghouse has picked the most viable concepts that will provide maximum accident tolerance as well as economics, and those that will make the most sense for utility adoption. All three options appear to be feasible from a technical and economic point of view. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Using probabilistic criteria in an assessment of the potential radiological consequences of the decommissioning of a nuclear research reactor

In order to assess the risk of radiological consequences of incidents and accidents in nuclear facilities it is important to contemplate their frequency of occurrence. It has to be shown that incidents and accidents occur sufficiently seldom according to their radiological consequences i. e. the occurrence frequency of radiological doses has to be limited. This is even demanded by the German radiation protection ordinance (StrlSchV), which says that in nuclear facilities other than nuclear power plants (NPP) in operation and for decommissioning, the occurrence frequency of incidents and accidents shall be contemplated in order to prove the design of safety measures and safety installations. Based on the ideas of the ICRP64, we developed a risk based assessment concept for nuclear facilities, which fulfils the requirements of the German regulations concerning dose limits in normal operation and design basis accidents. The general use of the concept is dedicated to nuclear facilities other than nuclear power plants (NPP) in operation and for decommissioning, where the regulation of risk assessment is less sophisticated. The concept specifies occurrence frequency limits for radiation exposure dose ranges, i. e. the occurrence frequency of incidents and accidents has to be limited according to their radiological effects. To apply this concept, scenarios of incidents and accidents are grouped in exposition classes according to their resulting potential effective dose to members of the general public. The occurrence frequencies of the incidents and accidents are summarized in each exposition class whereas the sum must not exceed the frequency limits mentioned above. In the following we introduce the application of this concept in the assessment of the potential radiological consequences of the decommissioning of a nuclear research reactor. We carried out this assessment for the licensing process of the decommissioning on behalf of German authorities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Prediction of number of breached rods following a LBLOCA of Candu plants using a BEPU approach

Radioactive doses following design basis accidents (DBA) have been important safety criteria of Candu nuclear power plant and they have been predicted in terms of the number of breached fuel rods. To support the licensing review on this concern, an analysis of LBLOCA has been conducted by using the BEPU method of KINS, KINS-REM. Number of Breached Rods (NBR) following a LBLOCA was predicted at 95 percentile probabilistic upper level in 95 percentile confidence level. Peak Cladding Temperatures (PCT) of the 84 bundles in the core pass 4 were calculated from the 124 MARS code runs in which the uncertainties of 10 major parameters including fuel thermal conductivity and break flow model were implemented. The fuel rod breaching criteria, PCT>1477 K, was used to determine the NBR 95/95. From the calculation, the predicted NBR 95/95 was 1591 rods and the calculated maximum NBR was lower than 2000 rods. Through the further improvements in feedback of the channel power behavior to thermalhydraulic calculation and in channel group modeling, NBR in more reliable level can be expected. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Characteristics Analysis of SGTR accidents under the Shutdown Conditions

The Steam Generator Tube Rupture (SGTR) accident is one of design basis accidents that could result in undesirable radiation release as it bypasses the containment. According to Ref. [1], SGTRs have been experienced in several nuclear power plants and Ulchin 4 of Korea is one of them. Basically, SGTR analysis assumes that it occurs at full power, the SGTR at Ulchin 4 in 2002, was in the shutdown condition. Therefore, the transient behavior was different from the cases in full power conditions and especially, the effectiveness of Emergency Operating Procedure (EOP) needs to be checked if it is also applicable to the shutdown conditions, because EOP is prepared for full power conditions. In this paper, we set up a MARS-KS code model for Ulchin 4 unit and simulated the SGTR accident for benchmarking the model to the actual plant behavior during the accident. And then, several case studies were done to investigate the consequence of SGTR and the effectiveness of the EOP, focusing on the amount of the primary to secondary discharge flow which was influenced by operator actions. At the end, we'll discuss about the EOP for the shutdown conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Multi-Scale Multi-physics Methods Development for the Calculation of Hot-Spots in the NGNP

Radioactive gaseous fission products are released out of the fuel element at a significantly higher rate when the fuel temperature exceeds 1600°C in high-temperature gas-cooled reactors (HTGRs). Therefore, it is of paramount importance to accurately predict the peak fuel temperature during all operational and design-basis accident conditions. The current methods used to predict the peak fuel temperature in HTGRs, such as the Next-Generation Nuclear Plant (NGNP), estimate the average fuel temperature in a computational mesh modeling hundreds of fuel pebbles or a fuel assembly in a pebble-bed reactor (PBR) or prismatic block type reactor (PMR), respectively. Experiments conducted in operating HTGRs indicate considerable uncertainty in the current methods and correlations used to predict actual temperatures. The objective of this project is to improve the accuracy in the prediction of local "hot" spots by developing multi-scale, multi-physics methods and implementing them within the framework of established codes used for NGNP analysis.The multi-scale approach which this project will implement begins with defining suitable scales for a physical and mathematical model and then deriving and applying the appropriate boundary conditions between scales. The macro scale is the greatest length that describes the entire reactor, whereas the meso scale models only a fuel block in a prismatic reactor and ten to hundreds of pebbles in a pebble bed reactor. The smallest scale is the micro scale--the level of a fuel kernel of the pebble in a PBR and fuel compact in a PMR--which needs to be resolved in order to calculate the peak temperature in a fuel kernel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Preliminary Assessment of the Dose near the Shield Plug at Fukushima Daiichi Unit 2 (1F2)

Since the accident at Fukushima Daiichi on March 11, 2011, caused by the Great East Japan earthquake and subsequent tsunami, there have been significant coordinated international efforts to better understand the accident progression, phenomenology and consequences. Of particular note is the Organisation for Economic Cooperation and Development Nuclear Energy Agency's (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF Project). Phase 1 of this ongoing effort focused on modeling the modeling of the three beyond design basis accidents at the site for the first six days of the accident. Phase 2 of the effort is a forensic effort and focuses on extending the accident simulations out to 21 days and assessing the consequences of the accident. This effort aims to inform the government of Japan and The Tokyo Electric Power Company (TEPCO)'s ongoing effort to decommission and decontaminate the site. One of the key questions posed by TEPCO in its efforts to decommission the site is the location of released radionuclides (RNs) that result in anomalously high dose rate measurements on and near the shield plugs of all three units. In this summary, we provide insight into the possible location of the source resulting in this high dose. Previous MELCOR accident progression analyses have predicted leakage of the containment through the drywell head, directly below the shield plug. Using this plausible accident scenario, MELCOR is able to predict the radionuclide source to the region below the shield plug and possible deposition amounts within the gaps of the shield plug. Using this predicted source and the transport and shielding code MCNP6 the dose in and around the shield plug of 1F2 was assessed. Unit 2 was chosen since the refueling bay is still intact and therefore a simpler initial problem compared to Units 1 and 3. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Enhancing workload assessments for validation activities associated with DBA and BDBA scenarios - 165

After the Fukushima Daiichi accident, nuclear regulators around the world have required that power reactor licensees develop more extensive emergency mitigating responses and severe accident management provisions beyond the defense-in-depth measures for design basis accidents (DBAs), previously in place. Workload assessments represent common validation techniques that are used to demonstrate that workers are able to perform tasks without unacceptable performance degradation. High workload is known to induced stress and fatigue and may severely diminish a worker's capacity to perceive, recognize, and respond appropriately during emergency or unanticipated events, which may result in undesirable consequences. In estimating workload during emergency and severe accident scenarios, power reactor licensees tend to rely on subjective measures of workload, such as NASA-TLX. Due to reported mismatches in the literature between subjective and physiologically-derived estimates of workload, it is prudent to see what more can be done to improve the current state of practice, in the context of emergency and severe accident conditions. To improve confidence in workload estimates, it is advocated that the nuclear industry integrate physiological-based measures into current practices by making use of 'on-body' or 'wearable' physiological sensors. In this paper, an overview of three different types of empirical workload approaches is provided. The advantages of wearable physiological sensors are considered in the context of extreme environments and occupations, with tangible examples including heat stress and pupillometry. Suggestions for a consensus forum on workload are provided and a research plan directed at improving the current practice of workload estimation is offered for consideration. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

A defense in depth approach for nuclear power plant accident management

An initiating event may lead to a severe accident if the plant safety functions have been challenged or operators do not follow the appropriate accident management procedures. Beyond design basis accidents are those corresponding to events of very low occurrence probability but such an accident may lead to significant consequences. The defense in depth approach is important to assure nuclear safety even in a severe accident. Plant Damage States (PDS) can be defined by the combination of the possible values for each of the PDS parameters which are showed on the nuclear power plant simulator. PDS is used to identify what the initiating event is, and can also give the information of safety system's status whether they are bypassed, inoperable or not. Initiating event and safety system's status are used in the construction of Containment Event Tree (CET) to determine containment failure modes by using probabilistic risk assessment (PRA) technique. Different initiating events will correspond to different CETs. With these CETs, the core melt frequency of an initiating event can be found. The use of Plant Damage States (PDS) is a symptom-oriented approach. On the other hand, the use of Containment Event Tree (CET) is an event-oriented approach. In this study, the Taiwan's fourth nuclear power plants, the Lungmen nuclear power station (LNPS), which is an advanced boiling water reactor (ABWR) with fully digitized instrumentation and control (I and C) system is chosen as the target plant. The LNPS full scope engineering simulator is used to generate the testing data for method development. The following common initiating events are considered in this study: loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), loss of offsite power (LOOP), station blackout (SBO). Studies have indicated that the combination of the symptom-oriented approach and the event-oriented approach can be helpful to find mitigation strategies and is useful for the accident management. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

The ATWS analysis of the HTR-10GT core induced by the earthquake - 18489

The 10 MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700 deg. C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750 deg. C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the Anticipated Transient Without Scram (ATWS) caused by an earthquake. The gravity acceleration of the earthquake will make the porosity of the pebble bed smaller, which will thus induce a positive reactivity. The packing fraction of the HTR-10GT core will increase to 0.64 from 0.61, while a reactivity of 1.24% will be inserted into an initial core and 0.788% for an equilibrium core. Then the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the reactor period becomes less than 20 seconds, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1223 deg. C. If a temperature coefficient error of 20% was considered during analyzing, the peak temperature of the fuel during the ATWS will reach 1266 deg. C. It was a little higher than 1230 deg. C --the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation has been raised to 1620 deg. C, so the HTR-10GT is safe during the ATWS caused by an earthquake. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Study on air ingress phenomena during a depressurization accident of a VHTR - 18632

A depressurization accident is the one of the design-basis accidents of a Very-High-Temperature Reactor (VHTR). When a depressurization accident occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, it is important to understand the mixing processes of different kinds of gases in the stable and unstable stratified fluid layers. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Therefore, in order to predict or analyze the air ingress phenomena during a depressurization accident, it is important to develop a method for the prevention of air ingress during an accident. We carried out experiments and numerical analysis using three-dimensional (3D) CFD code to obtain the mixing process of two-component gases and the flow characteristics of localized natural convection. This study also investigated a control method for the natural circulation of air through the injection of helium gas. The numerical model consists of a storage tank and a reverse U-shaped vertical rectangular passage. They are separated by a horizontal partition plate. One sidewall of the high-temperature side vertical passage is heated and the other sidewall is cooled. The low-temperature vertical passage is cooled by ambient air. The experimental procedures and numerical analysis process are as follows. Firstly, the storage tank is filled with heavy gas and the reverse U-shaped vertical passage is filled with a light gas. In the vertical passage of the high-temperature side, localized natural convection is generated by the temperature difference between the vertical walls. These flow characteristics were obtained from the experiments and steady-state analysis. The experiment and analysis began after the partition plate was opened. The results obtained from the experiments were quantitatively simulated using 3D numerical analysis. The two component gases were mixed via molecular diffusion and natural convection. After some time elapsed, natural circulation occurred through the reverse U-shaped vertical passage. These flow characteristics are the same as those of phenomena generated in the passage between a permanent reflector and a pressure vessel wall of the VHTR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

A systematic approach to the analysis of double faults within design basic accident methodology - 14501

Within the UK there is a general trend towards adopting Design Basis Accident (DBA) methodology for the assessment of radiological hazards, including criticality. DBA is a single fault methodology and comprises identifying the initiating event, its frequency and consequence. A fault with a very low initiating event frequency will not require further analysis. A fault sequence that terminates in a sub-critical configuration would also not require further analysis; although the identified passive features ensuring sub-criticality are subjected to engineering substantiation. Where faults do not pass this initial screening, two safety measures are generally required to ensure the fault does not progress to a criticality incident. This methodology provides a robust approach to single faults ensuring that an appropriate level and range of plant measures are identified and subjected to appropriate substantiation. However, a criticality is a cliff-edge effect and is dependent on multiple parameters. It is entirely possible under DBA to evaluate several single fault sequences and conclude that no measures are required because the single faults terminate in a sub-critical configuration. Importantly, for criticality safety, under DBA, there is no requirement to prevent important parameters from deviating from normal or upper bound 'safe' levels leaving the plant at greater risk in the event that further faults initiate. The combination of two faults could lead to a criticality incident but there is no current driver within DBA methodology to analyze these coincident or double faults. There is, therefore, potential to significantly under estimate the risk of criticality. This understanding is embodied within other methodologies such as the Double Contingency Principle. However, within the UK the DBA methodology has so significantly improved the analysis of single faults that, in some instances, consideration of double faults has taken a back seat. This is despite the UK regulator stating that 'criticality safety cases should employ the double contingency approach'. This paper presents a systematic approach to ensure that the potential for double faults or multiple parameter deviations to result in a criticality is effectively evaluated within the wider DBA methodology. The methodology determines which faults may be screened out from double fault analysis on the grounds of being 'unlikely' to initiate and being present for a 'short' duration. For those faults screened into double fault analysis, this guidance sets out the potential means of determining acceptable risk. These means include the specific evaluation of the risk of two criticality control parameters deviating simultaneously, the definition of additional safety measures and the definition of the boundary of safety for multiple parameters and the risk of exceeding this boundary. (authors)

42 ENGINEERING↗

MARVEL Hazard Evaluation ECAR-6440

MARVEL Hazards Evaluation evaluated the impacts of MARVEL operations, hazards, and postulated accidents. The hazard evaluation of MARVEL events and associated operations was performed for selection and evaluation of safety classification of systems, structures, and components (SSCs) and SSC safety functions, and for selection of design basis accidents (DBAs) applicable to the MARVEL microreactor design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

ITER CS Intermodule Support Structure

With five independently driven, bi-polarity power supplies, the modules of the ITER central solenoid (CS) can be energized in aligned or opposing field directions. This sets up the possibility for repelling modules, which indeed occurs, particularly between CS2L and CS3L around the End of Burn (EOB) time point. Light interface compression between these two modules at EOB and wide variations in these coil currents throughout the pulse produce a tendency for relative motion or slip. Ideally, the slip is purely radial as the modules breathe without any accumulative translational motion. In reality, however, asymmetries such as nonuniformity in intermodule friction, lateral loads from a plasma Vertical Disruption Event (VDE), magnetic forces from manufacturing and assembly tolerances, and earthquakes can all contribute to a combination of radial and lateral module motion. This paper presents 2D and 3D, nonlinear, ANSYS models which simulate these various asymmetries and determine the lateral forces which must be carried by the intermodule structure. Summing all of these asymmetric force contributions leads to a design-basis lateral load which is used in the design of various support concepts: the CS-CDR centering rings and a variation, the 2001 FDR baseline radial keys, and interlocking castles structures. Radial key-type intermodule structure interface slip and stresses are tracked through multiple 15 MA scenario current pulses to demonstrate stable motion following the first few cycles. Detractions and benefits of each candidate intermodule structure are discussed, leading to the simplest and most robust configuration which meets the design requirements: match-drilled radial holes and pin-shaped keys.

42 ENGINEERING↗

Liquid Salt Combined-Cycle Pilot Plant Design

The work described in this report is responsive to the Office of Fossil Energy program ‘Energy Storage for Fossil Power Generation.’ This Phase I report has been prepared by Pintail Power LLC, with support from Nexant ECA, Electric Power Research Institute (EPRI) and Southern Company Services as a deliverable for the U.S. Department of Energy for NETL Award DE-FE-00320016. The Liquid Salt Combined Cycle™ (LSCC™) technology provides large-scale energy storage integrated with Fossil Electric Generating Units (FEGUs) to meet critical needs in the energy transition by providing: • the lowest cost large-scale storage for time-shifting of renewable energy, • superior fuel efficiency to reduce GHGs from dispatchable resources, • flexible capacity and ramping to balance variability of wind and solar resources, • essential grid stability services to assure reliability of a low-carbon grid. The LSCC approach: • employs equipment that has already been proven in utility service, • uses safe, non-toxic, non-degrading, perpetual-life storage medium, • leverages and repurposes existing FEGU assets, • expands the value stack of energy storage to reduce market, financing, and commodity risks. Pintail Power has developed the LSCC technology to meet the need for reliable, efficient, and cost-effective integration of Variable Renewable Energy (VRE) into a low-carbon electric grid by coupling proven thermal energy storage with proven gas turbines, steam turbines, and heat transfer equipment. This novel approach is intended to address the key issues facing the grid and operators of renewable and fossil generating units including: • Overgeneration and curtailment of renewables, • Need for fast ramping dispatchable resources, • Improved efficiency and flexibility of fossil units, • Additional peaking capacity to support electrification of transportation and heating, • Provision of reliability services to support high penetration of VRE, especially synchronous inertia and fast frequency response. A Technology Readiness assessment by EPRI confirmed that LSCC technology consists of commercially proven hardware used in industrial and utility applications. Although the novel LSCC approach has not yet been demonstrated as a complete system, interfaces between major components have been conservatively specified. A Phase III pilot is planned to demonstrate equipment integration and operation. The patented innovation is removal of the evaporator section from the exhaust heat recovery system, with the evaporation performed by stored energy in a separate steam generator. This arrangement couples renewable and fossil power generation via long-duration energy storage to deliver cost, performance, and operational synergies, including superior charging and discharging flexibility, reduced fuel consumption and lower CO 2 emissions compared to conventional Combined Cycle Power Plants, and low-cost, large-scale energy storage. The LSCC technology is composed of proven equipment integrated with gas turbine exhaust heat in a novel system. During charging, electric heaters raise the salt temperature as it flows from the Cold Salt Tank to the Hot Salt Tank. During discharging, hot salt produces steam from feedwater that is heated with gas turbine exhaust, which also superheats steam to drive a steam turbine. LSCC technology can be added to any combustion-turbine to integrate renewable energy, provide needed grid services, and increase the value of fossil electric generating units based on the technology’s following attributes: • Long-duration storage enables time-shifting of VRE to avoid curtailment and impairment of renewable assets. • Long storage duration combined with fast-charging capability increases arbitrage opportunities by storing more energy when the price is low and discharging more hours when the price is high. • Long storage duration allows resource adequacy to be supplied across multiple days to increase reliability and reduce risk. • The stored energy reduces fuel heat rate and GHG emissions, and increases merit, so the LSCC dispatches earlier and longer to increase the plant’s capacity factor and asset value. • The stored energy enables pre-heating and startup of the steam cycle, without operating the gas turbine, to enable fast startup and ramping when dispatched for discharge. • The steam turbine can operate without the gas turbine so it can provide valuable synchronous inertia during charging without consuming fuel. • Fast frequency response and regulation services can be provided during charging using solid-state heater and pump controls to vary the charge power input in response to grid signals. • The LSCC system can be configured for resilience including black start, islanded/micro-grid operation, and even self-recharging of storage using either gas turbine power or gas turbine exhaust heat. The commercialization plan is to add LSCC technology to existing simple cycle gas turbine power plants with the 50MW GE LM6000 aero-derivative gas turbine as the reference design basis. A Techno-economic assessment of the reference design evaluated the benefits (Levelized Avoided Cost of Energy) and costs (Levelized Cost of Energy). The plant definition included all major systems and budgetary vendor quotes. Pintail Power and NexantECA developed the overall cost estimate for the LSCC plant up to the total plant cost level, following the DOE-NETL cost estimate guidelines at AACE Class 3 (-20%/+30%). This includes the equipment cost, bulk material, direct and indirect labor costs to arrive at the bare erected cost. Engineering costs are factored from the BEC and added to it to arrive at the EPC cost. Process and project contingencies were then factored from the EPC cost and rolled-up to yield the total plant cost of $\$$184 million for 1746 MWh of discharge electricity. • At $\$$105/kWh, the reference plant costs less than any of the Energy Storage Systems evaluated by PNNL in 2020 for the Energy Storage Grand Challenge. Operations and Maintenance cost estimates were scaled from combined cycle practice, assuming that the LSCC unit was co-located with and sharing some labor expense with other units, to arrive at $\$$2.2 million per year. Plant economics were evaluated using prices from the ERCOT Day-Ahead Market for calendar year 2019 (excluding the market disruptions from the COVID pandemic and the February 2020 deep freeze event). Assuming economic dispatch in the ERCOT Day-Ahead market, the reference plant capacity factor would have discharged for 2777 hours at 91.9 MW, a 31.66% capacity factor, with a marginal cost of $\$$25.59/MWh, and a LACE of $\$$82.41/MWh. Fixed charges were calculated according to EIA guidelines to arrive at an LCOE of $\$$83.48. The benefit-to-cost ratio of 0.99 suggests that the reference plant would have been cost-effective and competitive in the market. EPRI interviewed selected utilities to gauge the need for, applicability of and interest in the LSCC system. Several utilities are currently managing increased load growth along with the inclusion of increasing levels of renewable generation, putting pressure on conventional generation by requiring increased turndown requirements and ultimately lower capacity factors. All of the utilities interviewed have CO 2 reduction targets in the 2030-2050 timeframe that will severely limit the participation of fossil generation and require better utilization of carbon free generation. While there is limited opportunity for storage in the current markets, the utilities interviewed stated that there will be a substantial need for long duration energy storage in the future given the expected trends. Utilizing an energy storage system will generally be preferred over new gas capacity in some cases, with the capabilities of the LSCC system being a potential option for retrofit to existing simple cycle gas turbine units, allowing them to deliver greater participation in the market with lower carbon intensity. A technology gap assessment and technology maturation plan identified a pilot-scale demonstration as the final step before commercialization. Key gaps to be addressed during the Phase II FEED (Front-End Engineering Design) are component selection and design, commissioning procedures, and operational procedures and the control system for LSCC charging and discharging. The project team has been expanded to include Wood Group PLC as EPC. The proposed Phase II work leads to a pilot-scale engineering demonstration (TRL 6) to be conducted at Southern Company’s Plant Rowan, where the prototype system will perform “all the functions that will be required of the operational system.” The proposed pilot will facilitate commercialization (TRL-9) by scale-up to utility-scale systems integrated with peaking GTs or directly to facility scale systems using industrial GTs. The conceptual design for the pilot plant focuses on the novel integration aspects of LSCC technology. A slipstream of gas turbine exhaust will feed a waste heat recovery unit coupled to a molten salt steam generator heated by stored energy. The pilot is intended to demonstrate all key operating modes of the LSCC technology during charging, discharging and standby. The pilot equipment will be approximately one-seventh scale of the LM6000 commercial target and is expected to have commercial off-ramp potential for facility-scale applications.

01 COAL, LIGNITE, AND PEAT↗