The Radiation Protection Design of PWR Spent Fuel Dry Storage Facility

Radiation safety is an important part of safety assessment of spent fuel dry storage technology. This paper describes the radiation protection design of PWR spent fuel dry storage facility for radiation safety completed by China General Nuclear Power Corporation. Considering the special site conditions, Monte Carlo method is used to complete the precise calculation of the three-dimensional radiation dose field in the spent fuel storage building. Through the spent fuel storage module and the storage building with shielding function, radiation shielding design is completed to meet China’s regulatory requirements, which ensures radiation safety for workers and the public during the transport and storage of spent fuel. It will provide a reference for construction of spent fuel dry storage facility of CPR1000 and HPR1000.

Overview of the U.S. Department of Energy Advanced Waste Forms Development

In order to maintain the U.S. domestic nuclear capability, its scientific technical leadership, and to keep our options open for closing the nuclear fuel cycle, the Department of Energy, Office of Nuclear Energy (DOE-NE) invests in various R&D programs to identify and resolve technical challenges related to the sustainability of the nuclear fuel cycle. Sustainable fuel cycles are those that improve uranium resource utilization, maximize energy generation, minimize waste generation, improve safety and limit proliferation risk. DOE-NE chartered a Study on the evaluation and screening of nuclear fuel cycle options, to provide information about the potential benefits and challenges of nuclear fuel cycle options and to identify a relatively small number of promising fuel cycle options with the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the United States. The identification of these promising fuel cycles helps in focusing and strengthening the U.S. R&D investment needed to support the set of promising fuel cycle system options and nuclear material management approaches. DOE-NE is developing and evaluating advanced technologies for the immobilization of waste issued from aqueous and electrochemical recycling activities including off-gas treatment and advanced fuel fabrication. The long-term scope of waste form development and performance activities includes not only the development, demonstration, and technical maturation of advanced waste management concepts but also the development and parameterization of defensible models to predict the long-term performance of waste forms in geologic disposal. Along with the finding of the Evaluation and Screening Study will be presented the major research efforts that are underway for the development and demonstration of waste forms and processes including glass ceramic for high-level waste raffinate, alloy waste forms and glass ceramics composites for HLW from the electrochemical processing of fast reactor fuels, and high durability waste forms for radioiodine.

Modification of Filtered Air Intake Flowrate to Improve Control Room Habitability

Control room habitability (CRH) shall be maintained to provide adequate protection for control room operators, such that they can remain in the control room envelope (CRE) safely for an extended period and thus control the nuclear facility during normal and accident conditions. A critical objective of CRH systems is to limit operator doses and/or exposure to toxic gases. The CRH systems does this by the combination of the intake of filtered air, isolation of outside air, recirculation systems and etc. Among the parameters determining radioactivity in a control room (in proportion to radiation doses of operators), intake flowrate of filtered air is an important one. For different types of accident source terms, the evolution of operator doses in a control room versus intake flowrate were analyzed in this paper. It turns out that the increase of intake flowrate results in larger operator doses when inert radioactive gases are the dominant radioactive substances. On the contrary, increasing intake flowrate does good to lower the irradiation level of control room operators when radioactive aerosols dominate the source terms. The rationality behind this fact was interpreted in detail in this paper, with special attention paid to the unfiltered in-leakage rate. It can be inferred that an optimal intake flowrate probably exists leading to the minimum operator dose under an actual accident condition. This paper then performed a calculation analysis based on design parameters and source terms of design basis accident of LOCA (a large break loss of coolant accident) accident. The evolution of operator dose was found to be a U-curve versus increasing intake flowrate, which proved the existence of the abovementioned optimal intake flowrate of filtered air for CRH systems. Furthermore, the sensitivity analysis of intake flowrate was carried out to study the effects of unfiltered in-leakage rate and filtered recirculation. This study indicates that intake flowrate of filtered air can significantly influence the CRH. For different accidents, the intake flowrate should be properly modified rather than set as a fixed value. To optimize the radiological habitability of control rooms, the effects of unfiltered in-leakage must be taken into consideration. Besides, filtered recirculation is an effective way to control radiation exposure caused by iodine and radioactive aerosols.

Research on Deposition of Micro-Nano Aerosols in Rising Bubble Under Pool Scrubbing Condition

Aerosols as the main component of radioactive products in migration performance, which is an important factor that a unclear reactor accident present strong diffusion and affects the distributions of source and dose level in reactor containment, and they are therefore expected to be deposited in liquid phase such as in suspension pool and filtered containment venting device. In this paper, the deposition characteristics of micro-nano aerosols in rising bubble under pool scrubbing condition is studied with experiment, the aerosols size in the research range from 20 nm to 600 nm, and the bubble morphology mainly concern homogeneous bubbly flow. The results show that the deposition efficiency and mechanism of aerosol closely relate to gas flow rate, liquid level, particle size and bubbles size and so on. The aerosol deposition near 85nm is proved most difficult because of the convert of deposition mechanisms. In a high liquid level condition, micro-nano aerosol filtration efficiency is enhanced but gradually gradual. Under different gas flow rate, air bubble residence time and the bubble size distributions affect the filtration efficiency of aerosols.

Development of Laser Cleaning for Metallic Equipment

Laser cleaning study was performed on prepared samples using a nanosecond pulsed ytterbium fiber laser. To prepare samples, AISI 304L stainless steel samples were oxidized and implemented with non-radioactive contaminants in a controlled manner. In order to validate the cleaning process for metallic equipment polluted in nuclear installations, two types of contamination with europium (Eu) and with cobalt (Co) were studied. Eu was used as a simulator-product resulting from uranium fission, while Co — as an activation-product of nickel, which is a composing element of a primary coolant system of a reactor. The oxide layers have suffered laser scanning which was followed by the furnace treatment to obtain thicknesses in the range of 100 nm to 1 μm depending on the oxidation parameters [1] with a mean weight percentage of 1% of Eu and 1 % of Co in the volume of the oxide layer. Glow Discharge Optical Emission (GD-OES) and Mass Spectrometry (GD-MS) analyses have been performed to assess the efficiency of the cleaning treatment and to follow the distribution of residual contamination with a detection limit of 0.1mg/kg of Eu and Co. Decontamination rates up to 95.5 % were obtained. One of the identified reasons for this limitation is that the radionuclides are trapped in surface defects like micro cracks [2, 3]. Therefore, cleaning treatments have been applied on surface defects with controlled geometry and a micrometric aperture obtained by laser engraving and juxtaposition of polished sheets of AISI 304L stainless steel. The goal of this study is surface decontamination without either welding or inducing penetration of contamination into the cracks. GD-MS analysis and Scanning Electron Microscopy (SEM) were performed to analyze the efficiency of the treatment and the diffusion of contaminants in this complex geometry.

The Dose Constraint Calculation of High Radioactivity Level Waste Canister Surface

Radiation dose and personnel protection are among the safety goals of geological disposal of high-level radioactive waste. The calculation of the dose field on the surface of the packaging container is of great significance for the research on the dose constraint value of the repository. This paper built model consulting the Sweden KBS-3 canister, the temporal and spatial distribution of the dose rate on canister surface was calculated by Monte Carlo method, the temporal and spatial distribution of radiation dose rate of the tunnel was obtained. The research results showed that the photon dose rate on canister surface was greater than the neutron dose rate by 4 to 6 orders of magnitude, and the dose value of repository tunnel within 100 thousand years was lower than the ICRP recommended dose limit value (0.3 mSv/a) by 5 orders of magnitude.

Assessment of Radiological Source Term Releases for Potential Severe Accident Scenarios in a PWR SFP

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.

An Effect of Bismuth Ion on the Reduction of Terbium Ion in Molten LiCl-KCl Eutectic Salt

A salt waste generated from the pyroprocess contains residual actinides and needs to be purified for recycling of the salt and waste conditioning. A co-reduction process could be considered for removal of residual actinides from the salt waste, which contains lanthanides and residual actinides. In the study, specifically, an effect of Bi(III) ion on the electrochemical reaction of Tb(III) ion was investigated in the molten LiCl-KCl eutectic with BiCl3 and TbCl3 at 773 K using electrochemical techniques of cyclic voltammetry, square wave voltammetry and open circuit chronopotentiometry. Tb(III) has a single redox couple without Bi(III). However, the cyclic voltammograms obtained at tungsten electrode in LiCl-KCl-BiCl3-TbCl3 showed four redox couples. The square wave voltammogram in same condition also showed five reduction peaks. Cyclic voltammogram and square wave voltammogram was resolved to find the accurate peaks for redox reaction. Each peak indicates the formation of Tb-Bi intermetallic compound except Tb(III) reduction peak. From the phase diagram of Tb-Bi, it is inferred that each peak corresponds to TbBi2, TbBi, Tb4Bi3, and Tb5Bi3. The open circuit chronopotentiometry was conducted to estimate Gibbs free energy of formation of Tb-Bi intermetallic compound. The experimental results obtained from three kind of the electrochemical techniques showed that Tb-Bi intermetallic compounds were electrochemically formed under potential of Tb(III) reduction potential by co-reduction of Bi(III) and Tb(III). These results indicate that underpotential deposition by co-reduction could be used for Tb(III) removal from the salt waste with Bi(III).

Numerical Study of Water Droplet Heat Removal and Dynamics During its Impact Onto the Micro-Pillar Array at Elevated Temperature

The spray cooling and heat removal efficiency is one of the important aspect of nuclear thermalhydraulics and safety, especially for passive containment cooling after severe accidents. In order to design and optimize these systems effectively, computer modelling of the underlying mechanism of the liquid drop interaction with the hot solid surface would be necessary. Therefore, completeness, accuracy and reliability of the models that are being used in such sensitive areas are vital to the society and environment. Furthermore, the current powerful computer resources need to be fully exploited, so that the precision and the accuracy of the obtained computational results would be further enhanced. Nowadays, Volume-Of-Fluid (VOF) method is widely used in simulating the droplet dynamics, however these models provide estimations that are different in certain extents compare to the experimental results. In present work, we have used the level-set method to study the droplet dynamics and heat removal when the water droplet impact on the surface with different morphologies. The developed model which is based on the finite element method (FEM) has been benchmarked with previously performed experiments regarding the droplet bouncing on a flat hydrophobic surface; these estimations were in a good agreement with the previously published results. Moreover, hot solid surfaces with presence of micro-pillar has been considered to perform sensitivity study for different sizes of the micro-pillars and water droplets. In addition, it has been found that the heat transfer and droplet dynamic behavior would significantly vary in scenarios when the micro-pillars are presents in compare to a flat solid surface; it is observed that a better droplet spreading can be obtained with optimal size of micro-pillars that are present underneath of the droplet axial trajectory. The present study and the model would add valuable information to the field of heat transfer in aspect of spray cooling by investigating the feasibility of using the level-set method for a better estimation of fluid and heat transfer related results.

Buildup and Decay Analysis of Corrosion Products Activity in Primary Coolant Loop of AP-1000

The broad half-life range of Activated Corrosion Products (ACPs) results in major radiation exposure throughout reactor operation and shutdown. The movement of unpredicted activity hot spots in coolant loop can bring about huge financial and dosimetric impacts. The PWR operating experience depicts that activity released during reactor operation and shutdown cannot be estimated through a simple correlation. This paper seeks to analyze buildup and decay behavior of ACPs in primary coolant loop of AP-1000 under normal operation, power regulation and shutdown modes. The application of a well-tested mathematical model is extended in an in-house developed code CPA-AP1000, to simulate the behavior of dominant Corrosion Products (CPs), by programing in MATLAB. The MCNP code is used as a subroutine of the program to model the reactor core and execute energy dependent neutron flux calculations. It is observed that short-lived CPs (56Mn, 24Na) build up rapidly under normal operation mode and decay quickly after the reactor is shutdown. The long-lived CPs (59Fe, 60Co, 99Mo) have exhibited slow buildup under normal operating conditions and likewise sluggish decay after the shutdown. To analyze activity response during reactor control regime, operating power level is promptly decreased and in response specific activity of CPs also followed decreasing trend. It is noticed that activity of CPs drops slowly during reactor control regime in comparison to emergency scram. The results are helpful in estimating radiation exposure caused by ACPs during accessibility of the equipment in coolant loop, under normal operation, power regulation and shutdown modes. Moreover, current analyses provide baseline data for further investigations on ACPs in AP-1000, being a new reactor design.