STUDY OF THERMAL CREEP OF COATED CLADDING MATERIALS

Coated cladding materials are considered as a near-term concept for so-called Accident Tolerant Fuel materials. Their behavior in accidental conditions is mainly studied around the world, however, they need to survive in-reactor normal operation before reaching theoretical accidental conditions. An out-of-pile experiment was designed to study accelerated inward thermal creep and outward displacement of a cladding that occurs after pellet-cladding interaction. Four different materials deposited by the cold-spray technique were studied as well as the reference uncoated one. The results confirm theoretical prediction that due to a mismatch between the fundamental physical properties of each layer, high stresses will build up and the plastic strains expected will lead to coating cracking

Technology Selection for Offshore Underwater Small Modular Reactors

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead–bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.Naval Grou

On-Line Monitoring of Environment-Assisted Cracking in Nuclear Piping Using Array Probe Direct Current Potential Drop

A direct current potential drop method utilizing array probes with measurement ends maintaining an equalized potential designated as equi-potential switching array probe direct current potential drop (ESAP-DCPD) technique has been developed earlier at Seoul National University. This paper validates ESAP-DCPD technique by showing consistency among experimental measurements, analytical solution and numerical predictions using finite element analysis (FEA) of electric field changes with crack growth in metals. In order to examine its viability as an on-line monitoring of environment assisted crack growth at the inner surface of piping welds, artificial inner surface cracks were introduced in a full-scale weldment mockup pipe and stainless steel metal mockup pipe. The weldment was joined by low alloy steel and stainless steel pipes. The pipes were monitored by using ESAP-DCPD in laboratory environments. Optimization of electrical wiring configuration has produced results with significantly reduced noise for adequately long period of time. Then optimized experimental results were compared with the FEA prediction results for the mockup to show a good agreement. Also a round-robin measurement has been made at three laboratories. It has been found that the developed ESAP-DCPD can detect circumferential cracks with a depth of 40 % of wall thickness in stainless steel with a good detectability for further growth behaviors. For axial cracks, however, the measurements showed poor detectability. Hence the developed ESAP-DCPD system can be used to monitor large circumferential cracks that existing non-destructive examination techniques often fail to detect until leakage takes place.Korea (South). Ministry of Trade, Industry and Energy. Korea Institute of Energy Technology Evaluation and Plannin

Design of a Functionally Graded Composite for Service in High Temperature Lead and Lead-Bismuth Cooled Nuclear Reactors

A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required performance criteria of high strength and corrosion resistance. A Functionally Graded Composite (FGC) was created with layers engineered to perform these functions. F91 was chosen as the structural layer of the composite for its strength and radiation resistance. Fe-12Cr- 2Si, an alloy developed from previous work in the Fe-Cr-Si system, was chosen as the corrosion-resistant cladding layer because of its chemical similarity to F91 and its superior corrosion resistance in both oxidizing and reducing environments. Fe-12Cr-2Si experienced minimal corrosion due to its self-passivation in oxidizing and reducing environments. Extrapolated corrosion rates are below one micron per year at 700°C. Corrosion of F91 was faster, but predictable and manageable. Diffusion studies showed that 17 microns of the cladding layer will be diffusionally diluted during the three year life of fuel cladding. 33 microns must be accounted for during the sixty year life of coolant piping. 5 cm coolant piping and 6.35 mm fuel cladding were produced on a commercial scale by weld-overlaying Fe-12Cr-2Si onto F91 billets and co-extruding them, followed by pilgering. An ASME certified weld was performed followed by the prescribed quench-and-tempering heat treatment for F91. A minimal heat affected zone was observed, demonstrating field weldability. Finally, corrosion tests were performed on the fabricated FGC at 700°C after completely breaching the cladding in a small area to induce galvanic corrosion at the interface. None was observed. This FGC has significant impacts on LBE reactor design. The increases in outlet temperature and coolant velocity allow a large increase in power density, leading to either a smaller core for the same power rating or more power output for the same size core. This FGC represents an enabling technology for LBE cooled fast reactors

STUDY OF THERMAL CREEP OF COATED CLADDING MATERIALS

Coated cladding materials are considered as a near-term concept for so-called Accident Tolerant Fuel materials. Their behavior in accidental conditions is mainly studied around the world, however, they need to survive in-reactor normal operation before reaching theoretical accidental conditions. An out-of-pile experiment was designed to study accelerated inward thermal creep and outward displacement of a cladding that occurs after pellet-cladding interaction. Four different materials deposited by the cold-spray technique were studied as well as the reference uncoated one. The results confirm theoretical prediction that due to a mismatch between the fundamental physical properties of each layer, high stresses will build up and the plastic strains expected will lead to coating cracking. Keywords: Accident Tolerant Fuel; Fuel Cladding; Creep; CoatingUnited States. Department of Energy (Integrated Research Project Grant DENE0008416

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%

Influence of Aging on Corrosion Behaviour of the 6061 Cast Aluminium Alloy

The influence of AlFeSi and Mg2Si phases on corrosion behaviour of the cast 6061 aluminium alloy was investigated. Scanning Kelvin probe force microscopy (SKPFM), electron probe microanalysis (EPMA), and in situ observations by confocal laser scanning microscopy (CLSM) were used. It was found that Mg2Si phases were anodic relative to the matrix and dissolved preferentially without significantly affecting corrosion propagation. The AlFeSi phases’ influence on 6061 aluminium alloy local corrosion was greater than that of the Mg2Si phases. The corroded region width reached five times that of the AlFeSi phase, and the accelerating effect was terminated as the AlFeSi dissolved

MODELING THE PERFORMANCE OF HIGH BURNUP THORIA AND URANIA PWR FUEL

Fuel performance models have been developed to assess the performance of ThO[subscript 2]-UO[subscript 2] fuels that can be operated to a high burnup up to 80-100MWd/kgHM in current and future Light Water Reactors (LWRs). Among the various issues raised in high burnup fuel applications, the pellet rim effect, fission gas release (FGR), and response to reactivity initiated accidents (RIA) were of special interest in this work. These phenomena were modeled by modifying the NRC licensing codes FRAPCON-3 for normal operation and FRAP-T6 for transient conditions. These models were verified and compared to the results of previous thorium fuel studies and high burnup uranium fuel evaluations. The buildup of plutonium in the outer rim of LWR UO[subscript 2] pellets has been observed to create a region of high fuel burnup, fission gas buildup and high porosity at the fuel rim. The power distribution of the thoria and urania fuel was calculated using a neutronics code MOCUP. Due to the lower build-up of Pu-239 (less U-238 in ThO[subscript 2]-UO[subscript 2] fuel) and flatter distribution of U-233 (less resonance capture in Th-232), thoria fuel experiences a much flatter power distribution and thus has a less severe rim effect than UO[subscript 2] fuel. To model this effect properly, a new model, THUPS (Thoria-Urania Power Shape), was developed, benchmarked with MOCUP and adapted into FRAPCON-3. Additionally a porosity model for the rim region was introduced at high burnup to account for the larger fuel swelling and degradation of the thermal conductivity. The mechanisms of fission gas release in ThO[subscript 2]-UO[subscript 2] fuel have been found similar to those of UO[subscript 2] fuel. Therefore, the general formulations of the existing fission gas release models in FRAPCON-3 were retained. However, the gas diffusion coefficient in thoria was adjusted to a lower level to account for the smaller observed gas release fraction in the thoria-based fuel. To model accelerated fission gas release at high burnup properly, a new athermal fission gas release model was developed. Other modifications include the thoria fuel properties, fission gas production rate, and the corrosion model to treat advanced cladding materials. The modified version of FRAPCON-3 was calibrated using the measured fission gas release data from the Light Water Breeder Reactor (LWBR) program. Using the new model to calculate the gas release in typical PWR hot pins gives data that indicate that the ThO[subscript 2]-UO[subscript 2] fuel will have considerably lower fission gas release beyond a burnup of 50 MWd/kgHM. Investigation of the fuel response to an RIA included: (1) reviewing industry simulation tests to understand the mechanisms involved, (2) modifying FRAP-T6 code to simulate the RIA tests and investigate the key contributors to fuel failure (thermal expansion, gaseous swelling, cladding burst stress), and (3) assessing thoria and urania performance during RIA event in typical LWR situations. ThO[subscript 2]-UO[subscript 2] fuel has been found to have better performance than UO[subscript 2] fuel under RIA event conditions due to its lower thermal expansion and a flatter power distribution in the fuel pellet (less power and less fission gas in the rim region). Overall, thoria has been found to have better performance than urania in both normal and off-normal conditions. However, calculations using the modified FRAPCON-3 showed that the internal pressure and cladding corrosion at the required high burnup of 80- 100MWd/kgHM are not acceptable with the current fuel design. Therefore, advanced fuel designs (including larger gas plenum, larger fuel grains, advanced cladding materials), and carefully designed operating strategy (i.e. decreasing power history) were assessed and the results showed that the targeted high burnup can be achieved. Further investigation of burnup issues is needed, such as the distribution of hydrogen in the cladding for heterogeneous fuels, and response of high pressure fuel pins to a loss of coolant accident, in order to assure satisfactory high burnup behavior.Nuclear Energy Research Initiative (U.S.)United States. Dept. of Energy. Office of Nuclear Energ

The Effect of Weld Residual Stress on Life of Used Nuclear Fuel Dry Storage Canisters

With the elimination of Yucca Mountain as the long-term storage facility for spent nuclear fuel in the United States, a number of other storage options are being explored. Currently, used fuel is stored in dry-storage cask systems constructed of steel and concrete. It is likely that used fuel will continue to be stored at existing open-air storage sites for up to 100 years. This raises the possibility that the storage casks will be exposed to a salt-containing environment for the duration of their time in interim storage. Austenitic stainless steels, which are used to construct the canisters, are susceptible to stress corrosion cracking (SCC) in chloride-containing environments if a continuous aqueous film can be maintained on the surface and the material is under stress. Because steel sensitization in the canister welds is typically avoided by avoiding post-weld heat treatments, high residual stresses are present in the welds. While the environment history will play a key role in establishing the chemical conditions for cracking, weld residual stresses will have a strong influence on both crack initiation and propagation. It is often assumed for modeling purposes that weld residual stresses are tensile, high and constant through the weld. However, due to the strong dependence of crack growth rate on stress, this assumption may be overly conservative. In particular, the residual stresses become negative (compressive) at certain points in the weld. The ultimate goal of this research project is to develop a probabilistic model with quantified uncertainties for SCC failure in the dry storage casks. In this paper, the results of a study of the residual stresses, and their postulated effects on SCC behavior, in actual canister welds are presented. Progress on the development of the model is reported