Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “STEAM”

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 1,081 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Component level modeling of materials degradation for insights into operational flexibility of Existing Coal Power Plants

Increasingly, coal-fired power plants are required to balance power grids by compensating for the variable electricity supply from renewable energy sources. Fossil-fueled power plants, originally designed to be base loaded, will increasingly need to operate on a load following or cyclic basis. This demanding requirement for operational flexibility needs insights into accelerated material degradation arising due to the harsh operating conditions (e.g., fatigue, early oxide exfoliation due to stresses) along with current damage mechanisms (fireside corrosion, creep and erosion) observed in service. Our research objective is to develop component level modeling toolkit for materials-based degradation for two key mechanisms that can accelerate with cyclic operations. In more detail, this includes the fireside corrosion/steam oxidation/erosion/creep/fatigue of superheaters/reheaters and steam pipework and also the water droplet erosion/ fatigue of last stage steam turbine blades degradation mechanisms, that demand routine and sometimes unplanned maintenance and repair. The innovation is in developing a computational fluid dynamics/finite element (CFD/FE) modeling toolkit for the component level models of the boilers and low-pressure steam turbines in coal power plants that can tackle multidisciplinary failure mechanisms occurring concurrently for extreme environment materials. Lifetime assessment in such environments also needs to account for the unit-specific analyses, operational history and fuel feedstock; this can only be obtained by destructive analysis of components. This, in turn, enables validation of the model toolkits utilizing service feedback data, improving the probability of time/temperature dependent life prediction.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2023

Grid-Integrated Production of Fischer-Tropsch Synfuels from Nuclear Power

Idaho National Laboratory (INL) investigates the relative economic profitability of an integrated energy system (IES) coupling an NPP with a synfuel production process at selected case study locations across the United States. In the synfuel IES, a high-temperature steam electrolysis (HTSE) plant is thermally and electrically coupled with an NPP to produce zero-carbon hydrogen. The synthetic fuel is produced from this H 2 combined with a CO 2 supply using the reverse water gas shift process followed by the Fischer-Tropsch (FT) reaction. This analysis considers a system in which the CO 2 is sourced from regional CO 2 emitters via the construction and operation of pipeline-based CO 2 supply networks. Locating the FT plant at the same site as the NPP and HTSE plants enables the NPP to provide zero-carbon heat and power to the HTSE plant and zero-carbon power to the FT plant as well as avoid the requirement for long-distance H 2 product transport from the HTSE plant to the FT plant. Hydrogen storage is used to enable the NPP to dispatch power to the electrical grid (instead of the HTSE plant) when grid demand increases, thus enabling the FT plant to continue to operate in a steady-state production mode. The ability to cease hydrogen production for several hours within each day enables the NPP to provide power to the grid to balance the electricity market during peak periods and maximize revenues for the nuclear synfuel IES. The FT process design considered has a 99% carbon conversion efficiency. The use of nuclear energy and nuclear energy-derived hydrogen enables synfuel production to achieve this high level of carbon utilization. Additionally, the life-cycle carbon emissions of the nuclear-based synfuel production process are very low, with WTW emissions of approximately 25 gCO 2 e/MJ, including steam credits (generated from FT process excess heat), and approximately 7 gCO 2 e/MJ, if steam credits are excluded. This compares favorably with the WTW emissions of 90.5 gCO 2 e/MJ for a compression-ignition, direct injection (CIDI) vehicle with a fuel economy of 31.6 miles per gallon gasoline equivalent (MPGGE), using low-sulfur diesel produced using conventional petroleum production and refining processes. Several NPPs in various regions of the U.S. are considered as case study analyses. Supply locations and transportation via pipeline of the CO 2 feedstock to the NPP site are analyzed through the National Energy Technology Laboratory (NETL) CO 2 Transport Cost model. The team finds that the amount of CO 2 generated by different sectors is sufficient for the synfuel production process at all locations considered. The CO 2 transportation costs are functions of the distance of the source to the NPP location, the CO 2 capture cost at the source, and the quantity of CO 2 transported. Historical electricity prices for the NPP case study locations are collected and analyzed. Monthly average prices, price range, and duration of negative-price periods vary among these locations. For each location, an auto-regressive moving average (ARMA) model is trained on historical electricity price data. ARMA validation is done to ensure the synthetic price distributions represent one of historical prices with high fidelity. Synthetic time series from these ARMA models are used in a coupled dispatch and system optimization in the Holistic Energy Resource Optimization Network (HERON) to compute the differential net present value (NPV) of the IES. The team finds that this econometric is positive, ranging from $14M–1.3bn (2020) depending on the location. The optimal synfuel IES configuration to obtain this increase in NPV often maximizes the size of the synfuel production process with regards to the size of the NPP. However, the team shows that the NPP still plays a stabilizing role for the grid: In periods of high prices and high loads, more electricity from the NPP is sent to the grid. A high variability of electricity prices and extreme maximum prices tend to drive up electricity production. While it requires significant investment, the synfuel IES could increase the economic profitability for the existing fleet of LWRs across the country while still maintaining the grid stabilizer role of NPPs. During its lifetime, the main costs for the nuclear synfuel IES are the carbon feedstock transportation costs, followed by the capital expenses (CAPEX) and operation and maintenance (O&M) costs while the revenue comes first from the IRA H 2 production tax credit (PTC) and then from the sales of synfuel products. The profitability of the synfuel IES is most sensitive to the value of the hydrogen PTC and the synfuel products as well as the cost of the carbon feedstock, highlighting the importance of governmental incentives regarding hydrogen, carbon emissions, and synfuel in driving the deployment of future nuclear synfuel IESs.

08 HYDROGEN↗
DOE OSTI2024

Hydrogen Storage for Load-Following and Clean Power: Duct-firing of Hydrogen to Improve the Capacity Factor of NGCC Plants (Final Report, Phase II Pre-Front End Engineering Design Study)

GTI Energy (GTI) and team members Southern Company Services (SCS), Pacific Gas & Electric (PG&E) and the Electric Power Research Institute (EPRI) performed a Phase II Pre-Feed Study under contract DE-FE0032008 for Hydrogen Storage for Load-Following and Clean Power. The configuration of the proposed system consists of subsystems for on-site H 2 production, on-site H 2 storage (up to 54 MWth in commercial vessels) and H 2 combustion in a duct burner in a Heat Recovery Steam Generator (HRSG) integrated with an existing fossil asset. Here, the firing rate of the duct burner is varied to let the plant respond to fluctuations of electrical load and H 2 production is relatively constant by storing H 2 . The proposed system is an improvement over alternate low carbon dispatchable power options. Technoeconomic analyses show H 2 produced with GTI’s patented Compact Hydrogen Generator (CHG) with inherent carbon capture will be lower cost (Levelized Cost of Hydrogen, LCOH) relative to hydrogen produced with a Steam Methane Reformer (SMR) with an amine system for carbon capture. This lower cost hydrogen enables our integrated system to deliver electricity (Levelized Cost of Electricity, LCOE) at 17.4% lower cost relative to a SMR sourced H 2 -fired HRSG (with an amine system for carbon capture) – Steam Turbine Generator (STG). Our analysis using EPRI’s US REGEN macroeconomic model shows our system will have significant demand in the power market and therefore require significant capacity expansion (CHG plants built to deliver hydrogen) to deliver low cost, low carbon power. In Phase II, our Team has completed a detailed system definition including the development of process models, definition of battery limits, Process Flow Diagrams (PFDs), Piping and Instrumentation Drawings (P&IDs), and a plant layout. A preliminary design for the key components of the CHG was developed including component lists & specifications. An evaluation of environmental and permitting considerations was completed and included the development of an Environmental Information Volume (EIV). The duct burners, which are flexible and can burn hydrogen and/or natural gas, were defined and initial CFD analyses were completed.

03 NATURAL GAS↗
DOE OSTI2024

Development of Stable Solid Oxide Electrolysis Cells for Low-Cost Hydrogen Production

The project objective was to demonstrate a solid oxide cell-based steam electrolysis stack that exhibits robustness, reliability, endurance, hydrogen purity, and produces hydrogen at elevated pressure of 2 to 3 bar. Innovative materials and processing methods were evaluated to improve degradation characteristics. Performance improvement focused on nearly all layers involved in the cell and stack assembly. Primary attention was paid to zirconia-ceria interface resistance control via sintering optimization and decrease in degradation from the oxygen electrode by evaluating low strontium (Sr) or Sr-free composition for both the oxygen electrode and current collection layer. Stack robustness was addressed by validating redox tolerance of fuel electrode, confirming capability of cells to survive repeated thermal cycles, studying the effect of pressure on performance and degradation, evaluating the effect of contamination on fuel and oxygen electrode performance and degradation, and identifying mitigation strategies to improve performance. The characterization included evaluation of electrochemical performance and stability followed by microstructural analysis. At the cell level, performance and stability improvements were achieved by incorporating a Sr-free oxygen electrode and a denser oxygen electrode barrier layer. At the stack level, pressurized operation reduces demand on first stage compression, the redox tolerant fuel electrode mitigates risk from service interruptions, and improvements to interconnect coating alleviate chromium (Cr) contamination effects. The denser barrier layer was achieved by adding a sintering aid to the samaria-doped ceria (SDC) composition that reduced sintering temperature by 150 °C. The resulting density was on par with the baseline SDC barrier layer density and the lower sintering temperature resulted in less resistive phase formation during sintering. Button cell tests did not demonstrate a change in performance when exposed to silicon (Si) or manganese (Mn) impurities to the fuel electrode and Cr impurity to the oxygen electrode. More detailed study however is warranted. The project addressed SOEC performance and stability at the cell and stack levels through a systematic approach to known sources of degradation that were combined and tested in three stack tests using an electrolyte supported cell design to allow for evaluation of a variety of fuel and oxygen electrode compositions. STK-82 and STK-83 had identical compositions. STK-100 incorporated the best materials and processing variables developed under this and concurrent projects, and was tested at elevated pressure in steam electrolysis. • STK-82 recovered performance after redox and thermal cycling, demonstrating the robustness of the stack and seals. It exhibited stable performance in testing for 500 hours in SOEC mode, followed by 300 hours of cycling between SOEC and SOFC tests. Degradation during SOEC operation was 1.8 %/ 1,000 hours. • STK-83 generated hydrogen at >80% steam conversion, and oxygen above 98.5 % purity during pressurized operation. Both hydrogen and oxygen were generated at 3 barg pressure without the use of a pressure vessel. In addition to balanced pressure, electrolysis operation at 1 bar differential pressure across anode and cathode was also demonstrated to substantial the robustness of the cell and seal. • STK-100 measured at initial ambient pressure conditions showed an area specific resistance of 1.1 ohm-cm 2 , and STK-83 had 1.3 ohm-cm 2 .

08 HYDROGEN↗
NASA NTRS2023

Estimates of Permeability of Water in Rare Earth Silicates Using the Modified Theory of Deal and Grove

The salient elements of the theory of oxide growth under an environmental barrier are reviewed. In addition, the results from steam oxidation tests recently conducted on environmental barrier coated specimens are reported. The oxide thickness between the coating and substrate are reported as a function of time. Also, the results from previously conducted steam oxidation tests are reviewed. An approach to estimate the permeability of water through the various coating systems using the oxide thickness versus time measurements and the theory of oxide growth is demonstrated. Although the steam oxidation tests may have been performed under disparate conditions, the approach allows one to rank the protection efficiency of the various coating systems against substrate oxidation on an equivalent basis.

environmental barrier coatings↗
NASA NTRSDate not supplied

Bosch Carbon Catalyst Regeneration in Water III: Experiments for and Related to Thermal Shock

Bosch technology has the potential to close the atmospheric revitalization loop by providing theoretically 100% regeneration of metabolic carbon dioxide into oxygen. The main obstacle to using Bosch technology is over-pressurization issues in the carbon formation reactor. Prior work on Bosch catalyst regeneration established that agitating carbon coated beads in water broke up pressure causing clogs and regenerated the catalyst surface. This process proved to be time consuming with experiments taking at least one hour to regenerate ten grams of carbon coated beads. In an attempt to shorten the processing time, thermally shocking the carbon coated beads prior to mechanical agitation was tested. Agitating the carbon coated beads after exposure to a saturated steam environment was also tested to ensure that the steam generated by thermally shocking the carbon coated beads would not cause adverse effects. The rationale behind thermally shocking the carbon coated beads is that the bead would pull away from the carbon shell due to a difference in the thermal expansion coefficients. The results of this project show that the processing time can be cut in half when compared to previous work, and that saturated steam has little to no effect on regeneration.

Bosch Process↗
DOE OSTI2012

Superheater Corrosion Produced By Biomass Fuels

About 90% of the world's bioenergy is produced by burning renewable biomass fuels. Low-cost biomass fuels such as agricultural wastes typically contain more alkali metals and chlorine than conventional fuels. Although the efficiency of a boiler's steam cycle can be increased by raising its maximum steam temperature, alkali metals and chlorine released in biofuel boilers cause accelerated corrosion and fouling at high superheater steam temperatures. Most alloys that resist high temperature corrosion protect themselves with a surface layer of Cr{sub 2}O{sub 3}. However, this Cr{sub 2}O{sub 3} can be fluxed away by reactions that form alkali chromates or volatilized as chromic acid. This paper reviews recent research on superheater corrosion mechanisms and superheater alloy performance in biomass boilers firing black liquor, biomass fuels, blends of biomass with fossil fuels and municipal waste.

09 BIOMASS FUELS↗
DOE OSTI2012

Metal Matrix Microencapsulated (M3) fuel neutronics performance in PWRs

Metal Matrix Microencapsulated (M3) fuel consists of TRISO or BISO coated fuel particles directly dispersed in a matrix of zirconium metal to form a solid rod (Fig. 1). In this integral fuel concept the cladding tube and the failure mechanisms associated with it have been eliminated. In this manner pellet-clad-interactions (PCI), thin tube failure due to oxidation and hydriding, and tube pressurization and burst will be absent. M3 fuel, given the high stiffness of the integral rod design, could as well improve grid-to-rod wear behavior. Overall M3 fuel, compared to existing fuel designs, is expected to provide greatly improved operational performance. Multiple barriers to fission product release (ceramic coating layers in the coated fuel particle and te metal matrix) and the high thermal conductivity zirconium alloy metal matrix contribute to the enhancement in fuel behavior. The discontinuous nature of fissile material encapsulated in coated particles provides additional assistance; for instance if the M3 fuel rod is snapped into multiple pieces, only the limited number of fuel particles at the failure cross section are susceptible to release fission products. This is in contrast to the conventional oxide fuel where the presence of a small opening in the cladding provides the pathway for release of the entire inventory of fission products from the fuel rod. While conventional metal fuels (e.g. U-Zr and U-Mo) are typically expected to experience large swelling under irradiation due to the high degree of damage from fission fragments and introduction of fission gas into the lattice, this is not the case for M3 fuels. The fissile portion of the fuel is contained within the coated particle where enough room is available to accommodate fission gases and kernel swelling. The zirconium metal matrix will not be exposed to fission products and its swelling is known to be very limited when exposed solely to neutrons. Under design basis RIA and LOCA, fuel performance will be superior to the conventional oxide fuel since PCMI and rod burst, respectively, have been mitigated. Under beyond design basis accident scenarios where the fuel is exposed to high temperature steam for prolonged periods, larger inventory of zirconium metal in the core could negatively affect the accident progression. A thin steam resistant layer (e.g. alumina forming alloy steel), integrated into the solid rod during fabrication by co-extrusion or hot-isostatic-pressing, offers the potential to provide additional fuel protection from steam interaction: blanketing under a range of boiling regimes and under severe accident conditions up to high temperatures well beyond what is currently possible in the conventional fuel. A crucial aspect to the viability of M3 fuel in light water reactors is the reduced heavy metal load compared to standard pellet fuel. This study evaluated the design requirements to operate a Pressurized Water Reactor (PWR) with M3 fuel in order to obtain fuel cycle length, reactivity coefficients, and power peaking factors comparable to that of standard fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

The Phenix ultimate natural convection test

The French sodium cooled fast reactor Phenix was shut down in 2009 after 35 years of operation. Before decommissioning, a final set of tests were performed by the CEA during 9 months. Several topics were involved such as thermal hydraulics, core physics and fuel behaviour. Among these ultimate experiments, two thermal hydraulic tests were performed: an asymmetrical test consisting in a trip of one secondary pump and a natural convection test in the primary circuit. Recognizing the unique opportunity offered by these Phenix ultimate tests, IAEA decided in 2007 to launch a Coordinated Research Project (CRP) devoted to benchmarking analyses with system codes on the Phenix natural convection test. One objective of the natural convection test in Phenix reactor is the assessment of the CATHARE system code for safety studies on future and advanced sodium cooled fast reactors. The aim of this paper is to describe this test, which was performed on June 22-23, 2009, and the associated benchmark specifications for the CRP work. The paper reminds briefly the Phenix reactor with the main physical parameters and the instrumentation used during the natural convection test. After that, the test scenario is described: - initial state at a power of 120 MWth, - test beginning resulting from a manual dry out of the two steam generators, - manual scram, - manual trip on the three primary pumps without back-up by pony motors, - setting and development of natural convection in the primary circuit, in a first phase without significant heat sink in the secondary circuits and in a second phase with significant heat sink in the secondary circuits, by opening the casing of steam generators to create an efficient heat sink, by air natural circulation in the steam generators casing. The benchmark case ends after this second phase, which corresponds to the experimental test duration of nearly 7 hours. The paper presents also the benchmark specifications data supplied by the CEA to all participants of this CRP in order to perform calculations (core, primary circuit, primary pumps, IHX, shutdown system, operating parameters, test scenario and real test conditions). Finally, main test results and analyses are presented including the evolution of the core and of the heat exchangers inlet and outlet temperatures, and some local temperature measurements. The natural convection has been easily set up in the pool type reactor Phenix with different boundary conditions at the secondary side, with or without heat sink. The data obtained during this unique test represent some very useful and precious results for the development of SFR in a wide range of thematic such as numerical methods dedicated to thermal-hydraulics safety analyses (system codes, CFD codes and coupling system and CFD codes) and instrumentation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Treatment of Irradiated Graphite from French Bugey Reactor - 13424

Beginning in 2009, in order to determine an alternative to direct disposal for decommissioned irradiated graphite from EDF's Bugey NPP, Studsvik and EDF began a test program to determine if graphite decontamination and destruction were practicable using Studsvik's thermal organic reduction (THOR) technology. The testing program focused primarily on the release of C-14, H-3, and Cl-36 and also monitored graphite mass loss. For said testing, a bench-scale steam reformer (BSSR) was constructed with the capability of flowing various compositions of gases at temperatures up to 1300 deg. C over uniformly sized particles of graphite for fixed amounts of time. The BSSR was followed by a condenser, thermal oxidizer, and NaOH bubbler system designed to capture H-3 and C-14. Also, in a separate series of testing, high concentration acid and peroxide solutions were used to soak the graphite and leach out and measure Cl-36. A series of gasification tests were performed to scope gas compositions and temperatures for graphite gasification using steam and oxygen. Results suggested higher temperature steam (1100 deg. C vs. 900 deg. C) yielded a practicable gasification rate but that lower temperature (900 deg. C) gasification was also a practicable treatment alternative if oxygen is fed into the process. A series of decontamination tests were performed to determine the release behavior of and extent to which C-14 and H-3 were released from graphite in a high temperature (900-1300 deg. C), low flow roasting gas environment. In general, testing determined that higher temperatures and longer roasting times were efficacious for releasing H-3 completely and the majority (80%) of C-14. Manipulating oxidizing and reducing gas environments was also found to limit graphite mass loss. A series of soaking tests was performed to measure the amount of Cl-36 in the samples of graphite before and after roasting in the BSSR. Similar to C-14 release, these soaking tests revealed that 70-80% Cl-36 is released during roasting tests. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Development of Ce-type FeCrAl-ODS ferritic steel to accident tolerant fuel for BWRs

A Ce-type FeCrAl-ODS ferritic steel has been examined as a candidate Fe-Cr-Al alloy for the fuel cladding in BWRs. The properties examined in this study were the corrosion resistance to coolant water, the resistance to the high temperature steam and the interactions with neighboring materials (UO{sub 2} and B{sub 4}C). For the corrosion test, the specimens were corroded in an autoclave with circulating pure water at 633 K and a pressure of approximately 20 MPa. In addition to the Ce-ODS, MA956 and SS316L alloys were also corroded as reference materials. For the steam oxidation test, the specimens were oxidized in an electric furnace at 1473 K for 100 h and 1573 K for 25 h under wet Ar gas flow. The materials reaction test with B{sub 4}C was carried out at 1573 K and 1673 K for 1 h. The test with UO{sub 2} was performed in a high purity He gas at 1573 K and 1673 K for 1 h. A disk of UO{sub 2} pellet was set on the Ce-ODS specimen and the reaction couple was held in the pure Zr box to prevent a surface oxidation by residual oxidizing gases. The cross sections of reaction couples were observed by an optical microscope (OM) and a secondary electron microscope (SEM) with energy-dispersive spectroscope (EDS) after the materials interaction tests. The results show that the Ce-type FeCrAl-ODS steel has demonstrated a good corrosion resistance to coolant water and high accident tolerances in the oxidation with high temperature steam and in the interactions with neighboring materials. Those favorable properties would imply that the oxide dispersion strengthened (ODS) technology can be a candidate option for further development of accident-tolerant fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fuel cladding post-quench LOCA embrittlement: mechanical test relevance

During a loss-of-coolant-accident (LOCA), the fuel cladding is submitted to a high temperature increase in a steam environment, along with an internal pressure increase which would eventually lead to ballooning and burst of the cladding. Steam could then penetrate through the burst opening and lead to inner surface oxidation and secondary-hydriding of the cladding. Post-quench cladding rupture can then occur either in two-sided oxidized area with wall thinning due to ballooning, or in single-sided oxidized area with large hydrogen content. Common tests involved in the evaluation of the post-quench ductility are Ring Compression Tests (RCT), Expansion Due to Compression Test (EDC) or Axial Tensile Tests (AT). Questions arise about these mechanical tests, as for instance, the influence on sample failure of the oxide layer thickness partitioning between inner and outer diameter. A systematic study was carried out to compare several mechanical tests on both single-sided and two-sided Zircaloy-4 oxidized samples. Five different mechanical tests were compared, including RCT: C-shaped ring Compression Test (CCT), Axial Tensile test (AT), Ring Tensile Test with well-defined machined gage section (RTT) and plain Ring Tensile Test (pRTT). All samples were taken from cladding tubes which had been oxidized in steam at 1200 Celsius degrees and then water-quenched. Mechanical tests were performed at room temperature and two to three samples were tested for each oxidation conditions and for every sample geometry to assess the reproducibility. Metallographic examinations were carried out to measure the ZrO{sub 2} and αZr(O) layer thicknesses and the ex-β/α(O) phase fraction. Mean hydrogen content measurements were also performed systematically on each sample. Results indicate that the different mechanical tests have not the same sensitivity to the oxide layer partitioning: RTT and pRTT load at failure seems to depend mainly on the cumulated oxide layer thickness rather than on its partitioning, whereas this parameter has an impact on CCT and AT test results. For all tensile tests, stresses at failure were derived from load-displacement curves. The failure of brittle samples was assumed to be governed by linear elastic fracture mechanics. Using conventional stress intensity factor solution, a value of the material fracture toughness controlling the sample failure was determined from experimental results and discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Assessment of Core Status of Fukushima Daiichi Nuclear Power Plant

There were some events and phenomena that occurred which have been deemed specific to the Fukushima Daiichi NPP. The primary issues were as follows: - Direct steam leakage from the reactor pressure vessels (RPVs) to the dry-wells through the gaskets of the safety relief valves and/or due to creep damage of the main steam line, and damage of the in-core monitor guide tubes due to buckling. - Direct corium leakage from the RPVs to the pedestal floors through the melt portions of the in-core monitor housings. - Alternative water injection by the fire trucks with large amount of leakages through the branch lines. - Partial load operation of the Reactor Core Isolation Cooling system and the High Pressure Coolant Injection system. - Gas leakage from the primary containment vessels to the reactor buildings due to over pressure than the design maximum value. - Thermal stratification in the suppression pool resulting in incomplete steam condensation in the pool. The accident progressions were analyzed with the newly developed physical models for the above issues. And also considering the qualitatively estimated core status based on the plant parameters measured after the accident, the current core status was assessed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Innovative Decommissioning Strategies for Large Components - 18535

Within decommissioning projects the dismantling of steam generators as well as of the reactor pressure vessel with its internals form important and time consuming tasks. In order to optimize these, the reactor pressure vessel will be segmented into larger pieces in-situ and the larger pieces will be further segmented in a post segmentation area in order to suit the dimensional and radiological requirements of the final waste package. A similar approach is used also for the dismantling of Steam Generators. NUKEM Technologies Engineering Services has combined the two principles 'Rip / Store / Ship' and 'In-situ segmentation' to firstly segment the steam generator in-situ and then pack and ship all resulting parts allowing decontamination within 20' container to an existing decontamination facility. The aim is to free release 90 - 95% of the material. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Effects of bypass flow and power change and deviation on performance of thermal mixing structure of HTR-PM - 18424

The temperature of bypass flow not flowing through the reactor core is obviously lower than that of the main coolant helium in a High Temperature Gas-cooled Reactor (HTGR). In addition, change and deviation of reactor power because of power regulation will introduce higher radial temperature difference of coolant helium from reactor to steam generator. In order to investigate the effects of bypass flow and power change and deviation on radial temperature difference of the helium flow entering the steam generator, model experiments and simulation calculations are proposed to evaluate the mixing performance of the thermal mixing structure at HTR-PM reactor core outlet. The experiments on Test Facility Hot Gas Mixing (TF-HGM) are carried out to observe the influences of bypass flow and reactor power deviation on mixing performance. In addition, some simulation calculations are conducted for the main mixing structure of the TF-HGM facility and the thermal mixing structure of HTR-PM. According to the results of experiments and simulation calculations, the maximum radial temperature difference at the outlet of the mixing structure fulfils the requirement by steam generator, less than 30 deg. C ({+-}15 deg. C), under all considered conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Experimental study on a simplified facility of HTGR reactor building response to depressurization accidental scenarios - 18516

The analysis and understanding of air ingress events is an important aspect in the design of The High Temperature Gas Cooled Reactor (HTGR) accident scenarios including Depressurized Loss of Forced Cooling (D-LOFC) events that allow for the possibility of air ingress into the Reactor Pressure Vessel (RPV) as the result of a break in the helium pressure boundary which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. In order to characterize the air ingress into the vented low pressure compartments (VLPC) of NGNP HTGR during hypothetical small break D-LOFC accidents experimentally, a 1/28 down scaled simplified reactor building model was developed. A non-dimensional similarity approach was employed for the scaling analysis of the experimental facility. Two experiments were conducted under similar experimental conditions, with 1.1 kg/h helium injection through the top of the reactor cavity, resulting in oxygen concentrations reaching to steady state within 4 minutes. Differences between the steady state oxygen concentrations results in the steam generator cavity of the two tests with the same experimental conditions were 0.51{+-}0.086% at the top of the steam generator cavity and 0.29{+-}0.11% at the bottom of the steam generator cavity. Difference in the pressure in the reactor between the two experiments were 0.09{+-}0.026 psig. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Control system design for a small pressurized water reactor - 141

Small Pressurized Water Reactors (SPWR) have unique designs from the commercial large Pressurized Water Reactors (PWRs). Integral design is adopted. There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. Strong interactions exist among different loops. The pressure vessel is large in volume and the coolant inventory is also in a large amount. The inertia of the coolant is large and it takes more time for the primary system to respond to disturbances. The steam generator adopts the once-through design and the water inventory is small. The steam generator is very sensitive to disturbances. These unique characteristics challenge the control system design of SPWR. Relap5 is used to model the SPWR. Steady-state simulation is performed to verify the correctness of the model. In the reactor power control system, both the reactor fission power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance for the control system. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow and the letdown flow is compensated to calculate the charging flow needed. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A High Temperature Gas Reactor Integrated Energy System Using Energy Storage as a Buffer for Hydrogen Production

Analysis of an integrated energy system producing hydrogen in a high temperature steam electrolysis system via steam provided by a sensible heat thermal energy storage system charged by a high temperature gas reactor. Preliminary results over a continuous discharge cycle and single charging cycle shows minimal reactor impact despite 25% electrical power load change. Hydrogen production is held constant with steam provided by the thermal energy storage system.

08 HYDROGEN↗