Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “CLaDS”

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 1,549 records · Page 86
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Fabrication of Minor Actinide MOX Fuel for the AFC-2C and -2D Irradiations

Minor actinide bearing mixed oxide (MAMOX) fuels of nominal composition (U{sub 0.75}Pu{sub 0.20}Np{sub 0.02}Am{sub 0.03}) O{sub 2-x} were fabricated at Los Alamos National Laboratory (LANL) for the AFC-2C and AFC-2D irradiations that were conducted by the US-DOE under the Advanced Fuel Cycle program. The primary goal of this irradiation test series was to assess the irradiation performance behavior of MA-MOX fuel under prototypic fast reactor conditions. This paper describes the results of the MA-MOX fuel fabrication and pre-irradiation characterization effort and includes discussion of specific fresh fuel properties that affect irradiation performance. The as-fabricated MA-MOX compositions were (U{sub 0.77}Pu{sub 0.19}Np{sub 0.03}Am{sub 0.02})O{sub 2-x} with x=0.01 and 0.04. A comparison of fuel to clad interaction and cladding corrosion for fuels with O:M of 1.99 and 1.96 will be of particular interest during post-irradiation examination which is underway at the INL and discussed in a companion paper presented in this forum. The average fuel density was 87% of theoretical. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Start-up of Silo 130 Waste Retrieval at La Hague: first Step Towards Reducing Legacy Inventory - 20020

On June 25, 2019, the first grapple of waste and debris was lifted from Silo 130, one of the most challenging legacy waste storage silos at La Hague. This was the achievement of more than ten years of studies and efforts focused on solving some very unique technical and safety challenges, and on designing an innovative, efficient solution. The waste is from de-cladding materials from past Gas-Cooled Reactor fuel reprocessing operations. It is an unsorted mixture of magnesium cladding pieces, graphite debris, and various other metallic waste types. These materials can be chemically reactive and therefore need specific dispositions for handling and storing. The building is more than 40 years old and does not meet modern standards. French safety authorities have required the silo to be emptied. After about three years of operation, the bulk of this waste will be removed and placed in safe storage, pending the selection of a final conditioning solution. Removal of bottom debris will then be possible. When finished, this operation will bring a conclusion to a significant legacy-related hazard on the site. The silo will then be available for final cleaning and de-activation. This paper will describe the technical, safety, and economical challenges, and the various decisions and optimizations that led to the definition of the process and technology to be implemented. Several innovative features were developed and implemented, in all sections of the project: project organization, progressive approach for the waste disposition strategy, light building design, customized and robust retrieval technology, processing technologies involving the most modern robotic and sorting techniques, storage method... Several of these first-of-a-kind innovative features have been developed and qualified specifically for this application, using the detailed qualification procedure specified for all La Hague projects. The facility was built in less than five years, overcoming various difficulties along the line. Commissioning is now finished with the combined efforts of the project, engineering, start-up, and operating teams assisted by dedicated suppliers. The facility is now ramping-up. The paper will also report on the commissioning and start-up of the facility and describe the first operational results. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Drying Wet Stored and Corroded Magnox Fuel for Interim Dry Storage - 20187

To investigate the feasibility of drying corroded Magnesium alloy clad nuclear fuel, the corrosion products of inactive and unirradiated Magnox simulant dry corrosion product and dried corroded Magnox sludge were characterised by X-ray diffraction and thermogravimetric analysis. XRD identified brucite in the dried corrosion product and both brucite and hydromagnesite in the sludge. TGA identified mass loss on dehydration of ∼9.1% up to 250 deg. C and 27.9% from dehydroxylation up to 440 deg. C, with total mass loss of 41.9% up to 800 deg. C for the dried corrosion product. The sludge TGA showed 8.3% mass loss up to 250 deg. C from dehydration, 7.5% up to 330 deg. C from dehydroxylation and 31.4% mass loss up to 47.2% for decarbonation, with total mass loss of 49.2% up to 800 deg. C. Water removal up to 1.8 g was performed by cold vacuum drying (40 deg. C-120 deg. C) on a ≅14.8 g (wet) corroded Magnox sample. The process was monitored by observing pressure, dew point, temperature and gas flow changes supported by measuring the sample mass loss as water is removed. From these tests, it was observed that the dried corroded Magnox displayed some hygroscopic capacity which has implications for water retention following vacuum drying and the length of exposure required to achieve the desired dryness. Whilst at higher temperatures the water was removed faster, the increased temperature also risks the fuel cladding undergoing unwanted chemical reactions with the residual water during the process. However, at lower temperatures the achievable levels of dryness were reduced. Therefore, from these observations this work has identified that there is a temperature balance that may be necessary to optimise the drying process with respect to allowing the greatest level of dryness whilst restricting unwanted chemical reactions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Augmented Monitoring and Condition Assessment Program for SNF Wet Storage Life Extension - 20489

Approximately 27 MTHM of spent nuclear fuel (SNF) owned and managed by the U.S. Department of Energy, Office of Environmental Management is stored in the L Basin at the Savannah River Site (SRS). This 'DOE SNF' is comprised of approximately 12,000 aluminum-clad, aluminum-based fuel assemblies (∼7 MTHM), and approximately 2000 non-aluminum fuel assemblies (∼20 MTHM). A program is in progress to perform Non-Destructive Examination (NDE) of the fuel and their storage containers to characterize the materials' condition, and to evaluate the effects of service to enable continued safe wet storage of the SNF. The Augmented Monitoring and Condition Assessment Program (AMCAP), a two-part program to develop and implement remote underwater Non-Destructive Examination of the aluminum SNF, and of the containers of the non-aluminum SNF, respectively, is aimed at the characterization and evaluation of corrosion degradation of aluminum fuel and container materials. The predominant design of the aluminum SNF (ASNF) stored in L Basin is the Materials Test Reactor (MTR) equivalent design, a plate fuel design. These SNF are stored in 5'' diameter tubes or 5'' x 5'' squares (called bundles) that are nominally 12' or up to 14' long. The SNF is de-bundled and inspected using a custom-designed MTR Fuel Inspection Table. The inspection table provides for indexed fuel positioning for a video camera examination with controlled lighting. A total of 10 of the SNF originating from foreign research and test reactors were selected for examination based on burnup, enrichment, and prior damage caused by service/storage history. A special inspection campaign of these 10 assemblies is in progress. The observed corrosion damage included minor to moderate attack from general corrosion, pitting, crevice, end grain and galvanic corrosion. Example results from the completed inspections are shown and discussed. The focus is a comparison of the as-received condition versus the as-found current storage condition that will serve to validate Water Chemistry and Corrosion Monitoring Programs. The non-aluminum SNF (NASNF) stored in L Basin are of diverse design that includes various geometries with claddings of stainless steel, Zircaloy, and Hastelloy. The fuel core materials include uranium alloys, oxides/mixed oxides, and carbides. These SNF materials, originating from early experimental and test reactors, are in various physical forms including single fuel elements and cut pieces. The fuel is stored in L Basin in various configurations including in direct bundled storage in aluminum tubes and in isolation cans that are in the bundles or in a larger over-size storage container (OSC). Concern with inside-out corrosion and the potential loss of configuration control and ability to handle the storage containers prompted the development of remote NDE methods that include visual and UT technologies to assess the condition of the containers. Candidate stored materials were selected for a special inspection campaign; the development of the NDE methods for the inspection for galvanic, crevice and sediment-induced corrosion (inside-out) are discussed. A summary of the SNF storage in L Basin at the SRS, and an overview of the AMCAP to enable continued safe storage of DOE SNF are described. (authors)

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

Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly

The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Bond Strength of Plasma Sprayed Zr Diffusion Barrier on U-Mo Fuels

The United States government is continually committed to strengthening nuclear security and nonproliferation for reducing the threat of radioactive materials acquisition by terrorists. To meet this important mission, the US Department of Energy and National Nuclear Security Agency's (NNSA) Office of Material Management and Minimization (M{sup 3}) Reactor Conversion Program (CONVERT) seeks to minimize the use of highly enriched uranium (HEU) in civilian applications through work with governments and facilities around the world to convert research reactors to the use of non-weapons-usable low enriched uranium (LEU) fuels. Each reactor converted not only eliminates the need for HEU at civilian sites, but also reduces the amount of HEU being manufactured, stored, and transported, where the nuclear material is most vulnerable. In instances where suitable LEU fuels do not exist for the conversion of particular reactors, CONVERT contributes to development of new LEU fuels. Currently, the CONVERT Program is committed to development of a high-density U-Mo fuel that will be able to convert high performance research reactors, while establishing an efficient and economic fabrication capability to manufacture the fuel. To that end, Los Alamos National Laboratory (LANL) has been tasked with development of a method to affix zirconium (Zr) metal fuel diffusion barrier via plasma spraying, which will prevent a reaction between the U-Mo fuel and the 6061 aluminum cladding. A successful diffusion barrier requires adequate Zr layer thickness, uniform coverage, sufficient bonding at the Zr/fuel and Zr/aluminum cladding interface, and maintenance of the original nuclear fuel properties i.e. physical dimensions, alloy composition and dispersion, crystallographic phase concentration, neutronics, etc. This work outlines development of the Zr plasma spray processing techniques, parameters affecting the diffusion barrier/nuclear fuel bond strength, and characterization of the Zr diffusion barrier. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermo-Mechanical Analysis of SBLOCA with Loss of One Safety Injection System Using SCDAP/RELAP5

Design Extension Condition (DEC) for nuclear power plants has been being deliberated to bring it within a safety regulation boundary, following the 2011 Fukushima Daiichi accident. Moreover, a new regulation on the LOCA accident analysis designated as 10CFR50.45(c) has been proposed based on a concern of the non-conservatism of current criterion neglecting fuel burn-up effect on the cladding embrittlement. Therefore, a more comprehensive and realistic simulation of fuel rod behavior is required to cope with the stringent environment under LOCA accidents. A SCDAP/RELAP5 system analysis code has been used mainly on the prediction of severe accident phenomena. It has a set of mechanistic models for the fuel deformation, which plays an important role particularly in the early phase progression of the fuel melting. However, limited work has been done focusing on the models only to the fuel failure estimation through the DBA or candidates of DEC. Therefore, some analyses were performed here with a burnup dependent core model to understand the capability of SCDAP/RELAP5 to predict the thermo-mechanical behavior of the fuel cladding such as ballooning, burst and external/internal oxidation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Computational Studies of Gas Phase Reactions of Zirconium Tetrachloride (ZrCl{sub 4}) and Metal Chloride Impurities

Zirconium-based cladding materials that surround used nuclear fuel (UNF) rods represent a significant fraction of the total amount of high level waste that must be disposed of in long term geological repositories. The decontamination, recovery and reuse of these materials would significantly reduce the high level waste associated with light water reactors and could provide a new source of hafnium free zirconium for the fabrication of new nuclear components. One part in the strategy to recycle zirconium from used cladding materials involves developing a computational modeling (DFT) protocol to understand the reactions between ZrCl{sub 4} and metal chloride impurities that may occur in the gas phase during sublimation. Several of the metal chloride species present are known to form Lewis acid/base adducts in the gas phase and development of purification protocols require investigation into the properties of these ZrCl{sub 4} contaminates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Post-Irradiation Examination for Conventional and Advanced Nuclear Fuel

During irradiation in a nuclear reactor, the fuel pellet experiences non-uniform fission rate and strong thermomechanical loads, which result in alterations and restructuring occurring non-homogeneously in the fuel volume. The irradiated fuel is characterized by a highly heterogeneous structure, with a correspondingly heterogeneous distribution of phases and properties. It is a big technical challenge for a hot cell facility to be able to cover effectively a broad variety of fuel types and compounds, characterized by different compositions, physico-chemical properties, geometries and configurations. The first challenge is to provide testing environments and procedures which do not alter or degrade the original irradiated fuel properties and are technically suitable for hot cell or shielded facility operation. For instance, non-oxide fuel compounds require protective atmosphere to avoid sample oxidation during storage, preparation and testing. Moreover, suitable adaptations of tools and measuring devices must be available to characterize fuel configurations which differ from the conventional cylindrical pellet-cladding geometries (e.g. spherical fuel elements for pebble bed HTR and TRISO coated fuel particles). Finally, adequate scientific methods of investigation must be available to measure properties and study the behaviour of highly heterogeneous irradiated fuel. In addition to basic techniques for non-destructive and destructive examination of nuclear fuel systems (e.g. covering fuel and cladding or fuel and coatings, or other configurations), 'in-depth' methods are applied to determine physical, thermomechanical and micro-analytical properties of specific phases/features present in irradiated fuel. This paper presents relevant trends and examples from fuel safety studies performed at JRC-Karlsruhe and in other laboratories which highlight the 'added value' obtainable by applying combined techniques, high resolution tools or 3D methods of analysis for the experimental 'multiscale' characterization of irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Comparison of Thermal Diffusivity Degradation of Irradiated Zirlo and SiC-SiC Composite

Zirlo (Zircaloy cladding material from Westinghouse) has been widely used as the fuel cladding material in Light Water Reactors because of its high melting point, low susceptibility to corrosion, good thermal and mechanical shock resistance, and low degradation of mechanical properties under irradiation. But as Zirlo reacts with water at high temperature, which can trigger serious accidents, efforts have been made to explore better candidates to replace Zirlo in the water-cooled reactors. SiC-SiC composite is considered as one of the most viable options due to its comparable mechanical performances and corrosion resistance. However, one of the most important properties, thermal diffusivities of both Zirlo and SiC-SiC composite, and their degradation after irradiation, are not well known yet. Ion-irradiated samples are used to study the thermal property degradation caused by irradiation instead of the neutron-irradiated ones due to the significantly lower radioactivity. The microstructural damage is estimated from the software simulation SRIM/TRIM based on the energy and fluence of the ion, and the target material, and represented by DPA (displacement by atom). Different from neutron-irradiated samples, the damaged layer of ion-irradiated samples is only a few to a few tens of micrometers thick, and the damage is depth dependent. It is impossible to directly measure the bulk property change, and a microscale measurement technique is necessary. Laser-based photothermal reflectance technique is one of the ideal methods to measure local thermal conductivity and diffusivity with high spatial resolution. The main principle of this technique is to heat the solid sample using an intensity-modulated laser, and scan the sample with a constant intensity continuous-wave (CW) laser to detect the thermal wave propagation. On bulk, isotropic material, if the heating laser intensity is modulated as a sine wave, thermal diffusivity of the sample can be derived from the phase lag of the thermal wave, scan distance and modulation frequency in a straight forward manner. By using more specific analytical models, thermal conductivity and diffusivity in complicated cases, such as layered structure materials, composite materials, and/or anisotropic materials, are measurable as well. In this work, thermal diffusivity of the light damage layer of the post-irradiation Zirlo and SiC-SiC composite are measured and compared. The Zirlo sample is irradiated by Ni{sup 9+} ion, and SiC-SiC composite material is irradiated by Si{sup 7+} ion. The peak DPA of both sample is similar as ∼10. For ion-irradiated materials, DPA is known to increase with depth before it reaches the peak, and then sharply drop to zero. In term of light damage layer in this manuscript, we mean by the layer with a DPA smaller than 5% maximum value. In this layer DPA increases slowly and smoothly with depth, and the structure and thus thermal properties are more spatially uniform. For the SiC-SiC composite, the measurements are separately conducted on fibers along radial or axial direction, and on the matrix. Our measurement results show that with similar microstructural damage, thermal diffusivity of matrix SiC reduces more significantly than that of Zirlo. For fiber SiC that has a much lower thermal diffusivity than the matrix SiC before irradiation, the post-irradiation effect on thermal diffusivity is not obvious.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A Theoretical Investigation of the Chemical Behaviour of Irradiated U-Mo/Mg Fuel

A significant international effort has been put forth to replace Highly Enriched Uranium (HEU) fuel used in research and test reactors with Low Enriched Uranium (LEU) fuel to reduce proliferation concerns. Furthermore, some fuel processing issues have been encountered with some LEU candidate fuel designs. In order to address both of the foregoing issues, new fuel designs must be developed. One of these designs has U-Mo particles dispersed in a Mg matrix (U-Mo/Mg) that is cladded in an Al-alloy. While characterization analyses and thermophysical measurements have been reported for fresh U-Mo/Mg fuel, the effects of irradiation are not yet well known nor are they understood. Given that earlier international campaigns that investigated U-Mo/Al as a candidate fuel design resulted in consistent failures due to the formation of an interaction layer between the U-Mo fuel particles and Al meat, it seems prudent to theorize and predict any phases that may form as a result of irradiation of U-Mo/Mg fuel. The formation of these phases and any interaction that they may have with the cladding may potentially affect the reliability of this fuel design. This short paper provides a preliminary theoretical framework for the prediction of minor phases that may appear in irradiated U-Mo/Mg research reactor fuel, which should be confirmed via postirradiation examination at a later time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of Interaction Layer and U-Mo Size on Pore Growth in U-Mo/Al Fuel

U-Mo/Al dispersion fuel, composed of U-Mo alloy fuel particles embedded in the aluminum (Al) matrix, has been developed to convert high-enriched uranium fuel to low-enriched uranium for high-performance research reactors. The fueled zone, referred to as 'fuel meat,' is clad with aluminum alloy. Fuel meat and cladding are metallurgically bonded during fabrication. Irradiated U-Mo/Al dispersion fuel has shown stable performance in low fission rate and burnup ranges. However, excessive fuel meat swelling or breakaway swelling, which is believed an incipient failure, was observed at higher fission rate and burnup conditions. This breakaway swelling is believed to be a result from the combination of accelerated fuel swelling and interconnection of large pores formed outside of the UMo particles. These large pores were frequently found, particularly, at the interface between the Al and the fuel-matrix interaction layer (IL) that formed by irradiation-enhanced interdiffusion of U-Mo and Al. Kim et al. developed a model for pore growth using a mechanical equilibrium between gas pressure in the pore, interfacial energies, and the external stress at the pore surface. In this model, IL provides a path way for fission gas to reach the pores and also generates tensile stress fields. Hence, suppressing IL growth by applying a diffusion barrier, like a coating of the U-Mo particles, appears to be an effective method to prevent pore formation. Another remedy to prevent pore formation is to use large U-Mo fuel particles. Using larger U-Mo particles was thought advantageous because it reduces fission gas release. In this study, the effect of IL growth and U-Mo particle size on pore formation is examined using the model. The advantage was quantified in terms of porosity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Study on Neutronics Calculation Method of SCWR with Annular Fuel Assembly

Supercritical Water-Cooled Reactor (SCWR) is one of Gen-IV reactor systems for its advantages over LWRs, such as high thermal efficiency, simplified reactor system and economics. Annular fuel was proposed by MIT to supplant conventional solid fuel and improve power density in LWRs. Annular fuel is cooled from both inside and outside of the fuel. Thus a lower fuel temperature and cladding surface temperature can be achieved with higher power density. SCWRs will benefit from this advantage because high cladding temperature (usually over 650 degree C) is one challenge of SCWRs. Using annular fuel in SCWRs is very promising. This is also because the inner hole of the annual fuel can provide extra neutron moderation. Supercritical water density is too low to provide enough moderation. Different analysis codes have been developed all over the world to carry out SCWR pre-conceptual core design. Yamaji et al. developed SRAC/ASMBURN/COREBN to study Japan version of SCWR. Monti et al. used ERANOS to analyze HPLWR (High Performance Light Water Reactor). WIMS-D4/CITATION was also used to calculate HPLWR by Tashakor et al.. These codes employ a two-step method used in PWRs calculation. Lattice/Assembly calculation is carried out as the first step. In the second step, few group constant of different fuel assembly types is then used in core calculation. MCNP had also been used to study SCWR by Watta et al.. Due to large computation time cost, it's used mostly to analyze fuel assembly or core performance rather than pre-conceptual design, in which much optimization is needed. In this research, a neutronics calculation code for SCWR is developed using two-step method. This code is verified by comparing with Monte-Carlo code. Fuel assembly homogenization method is studied to improve the code accuracy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Improved Safety Injection Flow Map Associated with Target RCS Depressurization to Maintain Core Coolability of OPR1000

In Korea, Severe Accident Management Guidance (SAMG) for Optimized Power Reactor (OPR1000) has been developed through the research results related to the WOG SAMG, in which the operation phase shifts from the Emergency Operation Procedure (EOP) to the SAMG when the Core Exit Temperature (CET) reaches 923 K. Among 7 mitigation strategies envisioned, the third management strategy, injecting coolant into Reactor Coolant System (RCS) by depressurizing the RCS is one of the key strategies to fortify the in-vessel retention capability. As an aid to such mitigation strategy, a safety injection (SI) flow map has been developed by Lee et al. for predicting required flow rate associated with target RCS depressurization to recover in-vessel coolability. The mechanistic SI flow map used the CET increase rate and core water level decrease rate obtained from MELCOR code simulation. It should be noted that during the severe accident, the total heat generated is sum of the decay heat and the oxidation heat. Thus, it is necessary to consider additional heat sources by oxidation reaction between cladding, structure, control rod, and hot steam prior to the injection of coolant. It was believed that the SI model estimated required flow rate rather conservatively by considering the expected amount decay heat and the accumulated heat in the core based on the RCS pressure. However, it was confirmed that the SI model developed by Lee et al. could not represent the total heat removal especially since start of oxidation. Under transient conditions, the heat sources accumulated in the heat structures (fuel, cladding, supporting structures) should be embarked in the model for accurate contribution of the total heat source to the required flow rate. Therefore, the objective of this study is to improve the previous SI flow map predicting the injection flow rate associated with target RCS depressurization to satisfy the in-vessel coolability during severe accidents. Additional terms were developed to reflect accurate heat accumulating and included in the existing model. Resulting SI flow map includes the target RCS depressurization, which enables discharge of the required flow rate by utilizing the flow rate-pressure curve with operation of two high pressure safety injections (HPSIs). Also recent MELCOR simulation result of a postulated steam generator tube rupture (SGTR) accident verified the accuracy of the improved map.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Dispersion flattened extremely high-birefringent kagome lattice elliptic core photonic crystal fiber in THz regime

A highly birefringent elliptic porous-core photonic crystal fiber with kagome lattice cladding has been presented for terahertz (THz) wave transmission. The slotted airholes inside the elliptic core effectively disrupt the symmetry of the porous-core which offers extremely high modal birefringence whereas the kagome cladding structure results in significantly low effective material loss (EML). The birefringence of the proposed structure is 9.73×10{sup −2} and the EML is 0.056 cm{sup −1} for y-polarized mode at an operating frequency of 1 THz. The suggested structure also provides very flat dispersion over a bandwidth of 300 GHz. It is highly expected that this elliptic core waveguide would be remarkably useful in numerous polarization maintaining THz applications.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2019

Graphene oxide coated side polished surface plasmon resonance optical fibre sensor with varying polishing losses

Four batches of side-polished SMF optical fibres with different polishing losses, − 0.5 dBm, − 5 dBm, − 10 dBm and − 15 dBm were coated with a graphene oxide layer of ~ 0.5 μm via drop casting. TE polarized light was propagated through the light during experiments and four different transformer oils with different water contents were dropped onto the sensor. The drop in the power of transmitted light was recorded. The responses of the sensors showed good linearity for all, but the least polished fibre showed the weakest response in terms of power drop towards change in concentration. The − 5 dBm fibre showed the best response per unit concentration change. Additional tests showed the increased response for a GO coated side polished sensor compared to an uncoated side polished sensor. Simulation using COMSOL was performed to show the effects on the electric field of a sensor with and without GO coating. Another set of simulations were performed to show the effects of reducing cladding thickness on the coupling of the electric field. Simulations show that the optimal cladding thickness is not the most polished fibre, coinciding with our results that − 15 dBm loss fibre did not show the most response per unit change in the analyte.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2018

A Metallurgical Investigation of the Direct Energy Deposition Surface Repair of Ferrous Alloys

Among additive manufacturing (AM) processes, the direct energy deposition (DED) by laser is explored to establish its applicability for the repair of ferrous alloys such as UNS G41400 low-alloy steel, UNS S41000 martensitic stainless steel, UNS S17400 precipitation-strengthened martensitic stainless steel, and UNS S32750 super-duplex stainless steel. Unlike plating, thermal spray, and conventional cladding weld, DED laser powder deposition offers potential advantages, e.g., thin deposits, limited dilutions, narrow heat-affected zones (HAZ), potentially improved surface properties. In this investigation, all AM deposits were completed with an IREPA CLAD™ system using a powder feed of UNS N06625, an alloy largely selected for its outstanding corrosion resistance. This investigation first addresses topological aspects of AM deposits (including visual imperfections) before focusing on changes in microstructure, microhardness, chemical composition across AM deposits and base materials. It has been established that dense, uniform, hard (~ 300 HVN), crack-free UNS N06625-compliant AM deposits of fine dendritic microstructures are reliably produced. However, except for the UNS S32750 steel, a significant martensitic hardening was observed in the HAZs of UNS G41400 (~ 650 HVN), UNS S41000 (~ 500 HVN), and UNS S17400 (~ 370 HVN). In summary, this investigation demonstrates that the DED laser repair of ferrous parts with UNS N06625 may restore damaged surfaces, but it also calls for cautions and complementary investigations for alloys experiencing a high HAZ hardening, for which industry standard recommendations are exceeded and lead to an increased risk of delayed cracking in corrosive environments.

36 MATERIALS SCIENCE↗