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.

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.

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.

Possibility of Different Fuel Cycles Usage in GT-MHR

The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis reindicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another.

Radiotoxicity of Actinides During Transmutation in Final Stage of Atomic Power

Characteristics of a transmutation mode in final stage of atomic power are analyzed. In this stage, transmutation of actinides accumulated in transmutation reactors is performed without feed by actinides from other reactors. The radiotoxicity during first 20 years of transmutation is caused mainly by 244Cm. During following period of time, 252Cf is main nuclide. Contribution of 246Cm and 250Cf is 5–7 times less than that of 252Cf. During 50 years of a transmutation, the total radiotoxicity falls by 50 times. Long-lived radiotoxicity decreases slowly. During the period between T = 50 years and T = 100 years, long-lived radiotoxicity falls by 3.7 times. For each following 50 years after this period, long-lived radiotoxicity falls by 3.2 times. These results corresponding to neutron flux density 1014 neutr/(cm2s) in transmutation reactor demonstrate that the final stage of a transmutation should be performed with use of high flux transmutation facilities which provide shorter time of transmutation.

Modeling of Porous Crud Layer Based on Effects of Diffusion and Water Radiolysis on Anomalous Porous Crud Deposition on Fuel Pin Surfaces in PWRs

Many studies have been undertaken to understand crud formation on the upper spans of fuel pin clad surfaces, which is called axial offset anomaly (AOA), is observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling. Separately, researchers have considered the effect of water radiolysis in the primary coolant of PWR. This study examines the effects of radiolysis of liquid water, which aggressively participate in general cladding corrosion and solutes within the primary coolant system, in the terms of pH, temperature, and Linear Energy Transfer (LET). It also discusses the effect of mass transfer, especially diffusion, on the concentration distribution of the radiolytic products, H2 and O2, in the porous crud layer. Finally it covers the effects of chemical reactions of boric acid (H3BO3), which has a negative impact on the mechanisms of water recombination with hydrogen, lithium hydroxide (LiOH), which has a negative effect on water decomposition, dissolved hydrogen (DH), and some trace impurities.