Effect of Loading Pattern on Thermal and Shielding Performance of a Spent Fuel Cask

Two key considerations that must be addressed in licensing spent fuel storage systems are peak cladding temperatures and surface dose rates. Generally, storage systems are approved for a uniform loading of a design basis fuel. This analytical study addresses the effect of non-uniform loading patterns on peak fuel temperatures and cask surface dose rates. Three radial power distributions are considered: uniform loading, hotter assemblies in the center of the cask, and hotter assemblies near the wall of the cask. This paper summarizes the results of an analytical study [1] in which it was shown that, for the same total heat load in a cask, peak fuel temperatures are reduced by loading hotter assemblies around the outside of the cask’s basket. It was also shown that loading the hotter assemblies around the outside of the cask results in modest increases in surface dose rates.

Modeling of MOX Fuel Pellet-Clad Interaction Using ABAQUS

Post-irradiation examination (PIE) has indicated an increase in the outer diameter of fuel pins being irradiated in the Advanced Test Reactor (ATR) for the MOX irradiation program. The diameter increase is the largest in the region between fuel pellets. The fuel pellet was modeled using PATRAN and the model was evaluated using ABAQUS, version 6.2. The results from the analysis indicate the non-uniform clad diameter is caused by interaction between the fuel pellet and the clad. The results also demonstrate that the interaction is not uniform over the pellet axial length, with the largest interaction occurring in the region of the pellet-pellet interface. Results were obtained for an axi-symmetric model and for a 1/8 pie shaped segment, using the coupled temperature-displacement solution technique.

Completion of Experimental Breeder Reactor-II Sodium Processing at Argonne National Laboratory

The Experimental Breeder Reactor - II (EBR-II) at Argonne National Laboratory - West (ANL-W) was shutdown in September 1994 as mandated by the United States Department of Energy. Located in eastern Idaho, this sodium-cooled reactor had been in service since 1964, and was a test facility for fuels development, materials irradiation, system and control theory tests, and hardware development. The EBR-II termination activities began in October 1994, with the reactor being maintained in an industrially and radiologically safe condition for decommissioning. With the shutdown of EBR-II, its sodium coolant became a waste necessitating its reaction to a disposal form. A Sodium Process Facility (SPF), designed to convert sodium to 50 wt% sodium hydroxide, existed at the ANL-W site, but had never been operated. The SPF was upgraded to current standards and codes, and then modified in 1998 to convert the sodium to 70 wt% sodium hydroxide, a substance that solidifies at 65°C (150°F) and is acceptable for burial as low level radioactive waste in Idaho. In December 1998, the SPF began operations. Working with sodium and highly concentrated sodium hydroxide presented some unique operating and maintenance conditions. Several lessons were learned throughout the operating period. Processing of the 330 m3 (87,000 gallons) of EBR-II primary sodium, 50 m3 (13,000 gallons) of EBR-II secondary sodium, and 290 m3 (77,000 gallons) of Fermi-1 primary sodium was successfully completed in March 2001, ahead of schedule and within budget.

Radiation Induced Surface Activity Phenomenon: 2nd Report — Radiation Induced Boiling Enhancement

To delineate the effect of Radiation Induced Surface Activity (RISA) on boiling phenomenon, surface wettability in high-temperature environment or Leidenfrost condition and critical heat flux (CHF) of oxide metals irradiated by gamma rays were investigated. When the temperature of the heating surface reaches the wetting limit temperature, water-solid contact vanishes because of a stable vapor film between the droplet and the metal surface, i.e., a Leidenfrost condition. The wetting limit temperature increased with integrated irradiation dose. The CHF of oxidized titanium was improved up to 100% after 800 kGy 60Co gamma ray irradiated. Radiation Induced Boiling Enhancement (RIBE) phenomenon was firstly confirmed through the experiments.

Uncertainty Analysis of RBMK-Related Experimental Data

An attempt to validate state-of-the-art thermal hydraulic code ATHLET (GRS, Germany) on the basis of E-108 test facility was made. Originally this code was developed and validated for different type reactors than RBMK. Since state-of-art thermal hydraulic codes are widely used for simulation of RBMK reactors, further codes’ implementation and validation is required. The phenomena associated with channel type flow instabilities and CHF were found to be an important step in the frame of the overall effort of state-of-the-art validation and application for RBMK reactors. In the paper one-channel approach analysis is presented. Thus, the oscillatory behaviour of the system was not detected. The results show dependence on the nodalisation used in the heated channels, initial and boundary conditions and code selected models. It is shown that the code is able to predict a sudden heat structure temperature excursion, when critical heat flux is approached. GRS developed uncertainty and sensitivity methodology was employed in the analysis.

Thoria-Based Cermet Nuclear Fuel: Neutronics Fuel Design and Fuel Cycle Analysis

Cermet nuclear fuels have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle provide intriguing options for using thoria based cermet nuclear fuel in advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) Develop enabling technologies required for the economic application of this new fuel form. The following paper will first describe the fuel thermal performance model developed for the analysis of dispersion metal matrix fuels. The model will then be applied to the design and analysis of thorium/uranium/zirconium metal matrix fuel pins for light water reactors using neutronic simulation methods.

The Effect of the Hydrogen to Heavy Metal Ratio (H/HM) on Reactivity and Discharge Isotopics of Homogeneous Thoria-Urania Fuel

Calculations were performed using MOCUP, which includes the use of MCNP for neutron transport and ORIGEN for depletion. The MOCUP calculations were done using a unit cell (pin cell) model, where the ThO2 varied from 65–75wt% and the UO2 varied from 25–35wt%. The fission products and actinides being tracked in the calculations account for >97% of the parasitic captures in the fuel. The fuel pin was surrounded by four reflecting planes, where typical parameters were used for a 17×17 PWR assembly. The hydrogen to heavy metal ratio (H/HM) was varied by increasing or decreasing the water density in the cell. The results show that the drier lattices have insufficient reactivity due to the limited enrichment of the uranium. However, a slightly wetter lattice will increase the reactivity-limited burnup by 26% for the 25% UO2 – 75% ThO2, and 19% for the 35% UO2 – 65% ThO2 as compared to the standard coolant density. This is appears to be consistent with similar studies done with all-uranium lattices, where advantages are gained by hardening or further softening the neutron spectrum. Although some advantage is gained by softening the spectrum, the same can be said of all-uranium fueled cores. The spectral changes and, to a lesser extent, competing resonances between the 238U and bred-in 233U appear to hamper advantages in the conversion of thorium in homogeneous fuel that might otherwise be gained by shifting the neutron spectrum. Physically separating the uranium and thorium (e.g., in micro-heterogeneous and seed-and-blanket arrangements) have been shown alleviate this problem. A change in moderation may further enhance the reactivity-limited burnup of these lattices, and will be the focus of future work.

Stress Corrosion Cracking and Non-Destructive Examination of Dissimilar Metal Welds and Alloy 600

The United States Nuclear Regulatory Commission (USNRC) has conducted research since 1977 in the areas of environmentally assisted cracking and assessment and reliability of non-destructive examination (NDE). Recent occurrences of cracking in Alloy 82/182 welds and Alloy 600 base metal at several domestic and overseas plants have raised several issues relating to both of these areas of NRC research. The occurrences of cracking were identified by the discovery of boric acid deposits resulting from through-wall cracking in the primary system pressure boundary. Analyses indicate that the cracking has occurred due to primary water stress corrosion cracking (PWSCC) in Alloy 82/182 welds. This cracking has occurred in two different locations: in hot leg nozzle-to-safe end welds and in control rod drive mechanism (CRDM) nozzle welds. The cracking associated with safe-end welds is important due to the potential for a large loss of reactor coolant inventory, and the cracking of CRDM nozzle base metal and welds, particularly circumferential cracking of CRDM nozzle base metal, is important due to the potential for a control rod to eject resulting in a loss of coolant accident. The industry response in the U.S. to this cracking is being coordinated through the Electric Power Research Institute’s Materials Reliability Project (EPRI-MRP) in a comprehensive, multifaceted effort. Although the industry program is addressing many of the issues raised by these cracking occurrences, confirmatory research is necessary for the staff to evaluate the work conducted by industry groups. Several issues requiring additional consideration regarding the generic implications of these isolated events have been identified. This paper will discuss the recent events of significant cracking in domestic and foreign plants, discuss the limitations of NDE in detecting SCC, identify deficiencies in information available in this area, discuss the USNRC approach to address these issues, and discuss the development of an international cooperative effort.

A Life-Cycle Risk-Informed Systems Structured Nuclear Code

Current American Society of Mechanical Engineers (ASME) nuclear codes and standards rely primarily on deterministic and mechanistic approaches to design. The design code is a separate volume from the code for inservice inspections and both are separate from the standards for operations and maintenance. The ASME code for inservice inspections and code for nuclear plant operations and maintenance have adopted risk-informed methodologies for inservice inspection, preventive maintenance, and repair and replacement decisions. The American Institute of Steel Construction and the American Concrete Institute have incorporated risk-informed probabilistic methodologies into their design codes. It is proposed that the ASME nuclear code should undergo a planned evolution that integrates the various nuclear codes and standards and adopts a risk-informed approach across a facility life-cycle — encompassing design, construction, operation, maintenance and closure.

Design of Material Strength Test in Lead-Bismuth Flow

Liquid lead and lead-bismuth have drawn the attention as one of the candidate coolants of the fast breeder reactors (FBRs), and the accelerator driven transmutation systems (ADSs). In order to use the coolant to the systems, the physical and chemical characteristics of the heavy metals are necessary. This plan has been proposed for the strength test of materials in the liquid metal surroundings. The lead-bismuth circulation loop with the strength test has been designed, and the strength test of candidate materials has been planned.