Investigation of PWR Hull with a View to Downgrade

Abstract

The cladding materials remaining after the reprocessing of nuclear fuel, generally called hulls, are classified as high‐level radioactive waste. They are usually packaged in a container for disposal after being compacted, melted, or solidified into a heterogeneous matrix. Efforts to fabricate a better waste form from an environmental perspective have failed due to the technical difficulties encountered in the chemical decontamination of cladding hulls. In the early 1990's, the accumulation of radiochemical data on hulls and the advent of new technology such as laser or plasma have made the decontamination of hulls a viable option.

This paper summarizes information regarding the radiochemical analysis of spent nuclear fuel hulls through a literature survey, including the characteristics of the hulls of 32,000 MWd/tU burn‐up and 15 years cooling of Korean pressurized water reactor. The reduction of the radioactivity by peeling off the inner surface of the hulls via laser technology was evaluated.     

Thermal Expansion of Simulated Spent PWR Fuel and Simulated DUPIC Fuel

Thermal expansions of simulated spent PWR fuel and simulated DUPIC fuel were studied using a dilatometer in the temperature range from 298 to 1900 K. The densities of simulated spent PWR fuel and simulated DUPIC fuel used in the measurement were 10.28 gcm^–3 (95.4% of TD) and 10.26 gcm^–3 (95.1% of TD), respectively. The linear thermal expansions of the simulated fuels are higher than that of UO₂, and the difference between these fuels and UO₂ increases progressively with temperature. However, the difference between simulated spent PWR fuel and simulated DUPIC fuel is extremely small, less than the experimental error. For the temperature range from 298 to 1900 K, simulated spent PWR fuel and simulated DUPIC fuel have the same average linear thermal expansion coefficient of 1.39×10^–5K^–1. As the temperature increases to 1900 K, the relative densities of simulated spent PWR fuel and simulated DUPIC fuel decrease to 93.8% of initial densities at 298 K.     

Improvement of the sinterability of thermally-treated UO2 powder by horizontal rotary ball milling

Horizontal rotary ball milling has been demonstrated to be a useful method for reducing the particle size of ceramic powder in remote operation in shielded hot cells. Techniques, equipment and operating parameters, such as milling media, media wear and rotor speed were investigated with Al₂O₃ powder to evaluate its performance prior to contamination with nuclear fuel material. The established operating parameters were then verified with UO₂ powder, which had been produced by a thermal process to make fuel pellets. The sintering of the milled UO₂ powder showed the higher sintered densities obtainable by the milling, and the milling process seemed to be an important factor in improving the powder characteristics. This article is based on a presentation made in the “Symposium on Nuclear Materials and Fuel 2000“, held at the Korea Atomic Energy Research Institute (KAERI), Taejon, Korea, August 24–25 under the auspices of the Ministry of Science and Technology (MOST).     

Fabrication of DUPIC fuel pellets using high burn-up spent PWR fuel

Technology for the direct usage of a spent PWR fuel in CANDU reactors (DUPIC) was developed in KAERI to reduce the amount of spent fuel. DUPIC fuel pellets were fabricated using a dry processing method to re-fabricate CANDU fuel from spent PWR fuel without any intentional separation of fissile materials or fission products. The DUPIC fuel element fabrication process satisfied a quality assurance program in accordance with the Canadian standard. For the DUPIC fuels with various fuel burn-ups between 27,300 and 65,000 MWd/tU, the sintered pellet density decreased with increasing fuel burn-ups. Fission gas releases and powder properties of the spent fuel also influenced the DUPIC fuel characteristics. Measurement of cesium content released from green pellets revealed that their sintered density significantly depended on sintering temperature history. It was useful to establish a DUPIC fuel fabrication technology in which a high-burn-up fuel with 65,000 MWd/tU was treated.     

Radioactive Waste Arisings from Various Fuel Cycle Options

This study examines all kinds of waste volumes from various fuel cycle options including DUPIC (Direct Use of Spent PWR Fuel In CANDU) fuel cycle and compares each other. The fuel cycle option considered the PWR (Pressurized Water Reactor) once-through cycle, the PHWR (Pressurized Heavy Water Reactor) once-through cycle and the thermal recycling option using an existing PWR with MOX (Mixed Oxide) fuel. This study focuses on the radioactive wastes including mill waste, low-level waste and high-level waste generated by all fuel cycle steps, which can be one of the effectiveness measures of waste management. All waste disposition volume is estimated in terms of m³/GWe-yr. We find in the estimation of radioactive waste volume that PWR-MOX option has the lowest mill tailings and spent fuel volumes among the options, but the option has high volume of ILW and HLW. Mill tailings and spent fuel volumes of the DUPIC fuel cycle are lower than those of other competitive options such as PWR-PHWR once-through cycle. PWR once-through cycle has the lowest LLW and ILW volume among the options, but has high mill tailings and spent fuel volume. The data obtained in this study would be helpful to further estimate environmental effect and/or waste disposition costs in various fuel cycle options.     

The thermal diffusivity of simulated DUPIC fuel for irradiation test

The thermal diffusivity of simulated DUPIC fuel with an equivalent burnup of 35,000 MWd/tU (3.73 at.%), which was to be tested for irradiation at the HANARO research reactor, was measured using the laser flash method in a temperature range from room temperature to 1623 K. The density of the simulated DUPIC fuel used in the measurement of thermal diffusivity was 10.16 g/cm³ (94.2% of theoretical density) at room temperature, and the diameter and thickness were 10 mm and 1 mm, respectively. The thermal diffusivity decreased from 0.0205 cm²/s at room temperature to 0.0060 cm²/s at 1623 K. The thermal diffusivity of simulated DUPIC fuel was lower than that of UO₂ by 36.8% at 300K and by 12.7% at 1623 K. The difference in thermal diffusivity between the simulated DUPIC fuel and UO₂ was large at room temperature, and decreased as the temperature increased. This article is based on a presentation made in the “Symposium on Nuclear Materials and Fuel 2000” held at the Korea Atomic Energy Research Institute (KAERI), Taejon, Korea, August 24–25 under the auspices of the Ministry of Science and Technology (MOST).     

Characterization of irradiated simulated DUPIC fuel

A simulated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel was irradiated at HANARO research reactor of KAERI in 1999. Post-irradiation examinations, such as measurements of γ-scanning, profilometry, density, hardness, microstructure, and fission product distribution were performed on the irradiated simulated DUPIC fuel. In γ-scanning, the intensity along the axial direction was sharply decreased at the areas between the pellets. There was no significant change in the profilometry of SEU-1.47%, but variation was detected in SEU-2.19%+F.P by 67 μm, and the peaks precisely coincided with the ridges of the pellets. The marked difference between SEU-1.47% and SEU-2.19%+F.P pellets after irradiation was the configuration of cracks arised in the pellets. Some large equiaxed grains of 11.1 μm were observed at the center of the SEU-2.19%+F.P pellet, while the grain size near the surface of the pellet was remained almost the same as the original grain size of 5.58 μm. The hardness had no tendency toward change to the direction, but average hardness was increased as much as 10% compared with a fresh simulated DUPIC fuel. This article is based on a presentation made in the “Symposium on Nuclear Materials and Fuel 2000”, held at the Korea Atomic Energy Research Institute (KAERI), Taejon, Korea, August 24–25 under the auspices of the Ministry of Science and Technology (MOST).     

Remote Fabrication of DUPIC Fuel Pellets in a Hot Cell under Quality Assurance Program

The Korea Atomic Energy Research Institute (KAERI) has been developing the Direct Use of Spent Pressurized Water Reactor (PWR) Fuel in the CANada Deuterium Uranium (CANDU) Reactors (DUPIC) fuel fabrication technology since 1992, and the basic DUPIC fuel fabrication process was established in 2002. In order to demonstrate the robustness of the DUPIC fuel fabrication process through the irradiation test, it is important that a Quality Assurance (QA) program should be in place before a fabrication of the DUPIC fuel. Therefore, the Quality Assurance Manual (QM) for the DUPIC fuel was developed on the basis of the Canadian standard, CAN3-Z299.2-85. This manual describes the quality management system applicable to the activities performed for the DUPIC fuel fabrication at KAERI. In order to demonstrate the DUPIC fuel fabrication technology and produce qualified DUPIC fuel pellets, the process qualification tests were performed, which include three pre-qualification tests and three qualification tests. The characteristics of the DUPIC fuel pellets such as the sintered density, grain size, and surface roughness were measured and evaluated in accordance with the QA procedures. The optimum fabrication process of the DUPIC fuel pellet was also established based on the qualification results. Finally a production campaign was carried out to fabricate the DUPIC fuel pellets at a batch size of 1 kg following the QA program. As a result of the production campaign, qualified DUPIC fuel pellets were successfully produced and, therefore, the remote fuel fabrication technology of the DUPIC fuel pellet was demonstrated.     

Trapping characteristics of cesium in off-gas stream using fly ash filter

The cesium trapping characteristics with changing reaction temperature, carrier gas and gas velocity by the fly ash filter were analyzed. The SEM (Scanning electron microscope) on the pore structure of the fly ash filter showed that pores up to 0.1 mm in diameter were widely interconnected with each other throughout the whole structure of the filter. According to the XRD (X-ray diffraction) analysis for the cesium compound trapped on the fly ash filter, the thermally stable pollucite phase was formed. The cesium trapping quantity of the fly ash filter was increased with increasing reaction temperature, whereas it was decreased with increasing gas velocity. SEM showed that the fly ash filter after trapping gaseous cesium had mullite phase of needle-like crystals and pollucite phase of bulky crystals with rough surface. Presented at the Int’l Symp. on Chem. Eng. (Cheju, Feb. 8-10, 2001), dedicated to Prof. H. S. Chun on the occasion of his retirement from Korea University.     

Fabrication of simulated DUPIC fuel

Simulated DUPIC fuel provides a convenient way to investigate the DUPIC fuel properties and behavior such as thermal conductivity, thermal expansion, fission gas release, leaching, and so on without the complications of handling radioactive materials. Several pellets simulating the composition and microstructure of DUPIC fuel are fabricated by resintering the powder, which was treated through OREOX process of simulated spent PWR fuel pellets, which had been prepared from a mixture of UO₂ and stable forms of constituent nuclides. The key issues for producing simulated pellets that replicate the phases and microstructure of irradiated fuel are to achieve a submicrometre dispersion during mixing and diffusional homogeneity during sintering. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent PWR fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent PWR fuel agrees well with the other studies. The leading structural features observed are as follows: rare earth and other oxides dissolved in the UO₂ matrix, small metallic precipitates distributed throughout the matrix, and a perovskite phase finely dispersed on grain boundaries.