A GoldSim Modeling Approach to Safety Assessment of an LILW Repository System

A program for the safety assessment and performance evaluation of a low- and intermediate level waste (LILW) repository system has been developed by utilizing GoldSim [1]. By utilizing this nuclide transport in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release are modeled and evaluated. To demonstrate its usability, some illustrative cases under the selected scenarios including the influence of degradation of manmade barriers, pumping well drilling, and the natural disruptive events such as a sudden formation of preferential flow pathway have been investigated and illustrated for a hypothetical LILW repository. Even though all the parameter values applied to a hypothetical repository are assumed without any real base, the illustrative cases could be informative especially when seeing the result of the probabilistic calculation or sensitivity studies with various scenarios that possibly happen for nuclide release and further transport.

Alpha Radioactivity Monitor Using Ionized Air Transportation for Large Size Uranium Waste: Part 1—Large Measurement Chamber and Evaluation of Detection Performance

We present an ionized air transportation type alpha radioactivity monitor to efficiently perform the clearance level inspection for large size uranium waste and its detection performance. In previous work, we developed a prototype monitor with an about 1000 mm cubic measurement chamber to measure the cut waste. However, in a survey of target waste, we found that it is desired to measure not only the cut waste but also the lengthy waste such as uncut cylinders. Therefore, we developed an alpha radioactivity monitor with a long and large measurement chamber (effective sizes: 500 mm x900 mm x3200 mm) for long and large cylindrically-shaped waste (maximum size: 300 mm in diameter and 3000 mm in length, weight: 10 to 200 kg). We aimed <1000 Bq as the target value of Alpha radioactivity Detection Limit (ADL), which is one-tenth of the clearance level (1 Bq/g) for 10 kg waste. The issue to size up the measurement chamber was to suppress the reduction in sensitivity of alpha radioactivity. To overcome this, we enhanced an air fan power and optimized an ion sensor design. Using this monitor, we measured and evaluated ADLs for several cases supposing the practical applications (long cylinders with a smooth surface, bump, or concavity and convexity, and pipes with several small diameters). The resulting ALD ranged from 60 Bq to 120 Bq and sufficiently satisfied the target ALD (<1000 Bq). In conclusion, this monitor has sufficient performance for the clearance level inspection for large uranium waste.

Improvement of Radioactivity Inventory Evaluation Procedure in Preparatory Tasks for Decommissioning

Preparatory tasks for decommissioning of nuclear power plant start with radiological characterization. Residual radioactivity inventory evaluation is a main part of the characterization. Reliable information on the inventory is important for specification for decommissioning plan. Japan Atomic Power Company (JAPC) has already started these tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1). We can optimize decommissioning plan using the information. To obtain the reliable information, we improved an evaluation procedure. The procedure is divided into two main steps. First step is neutron flux distribution calculation and second one is radioactivity distribution calculation. Radioactivity distribution is calculated using neutron flux distribution. In this work, we improved the evaluation procedure to obtain the reliable information on the inventory. Because of the limitation of computer resource, two-dimension (2D) approximation model was applied to radioactivity distribution around Reactor Pressure Vessel (RPV). We can calculate reliable 2D neutron flux distribution by having better understanding of neutron transport phenomena. Neutron flux was measured at 30 locations in TS-1 Primary Containment Vessel (PCV) using activation foils. And in order to understand the neutron transport phenomenon inside the PCV, we also calculated neutron flux distribution with the three-dimensional (3D) discrete ordinates method calculation (Sn) code. By consideration about the result of the measurement and 3D calculation, we could understand the characteristics of the neutron flux distribution inside the PCV. To simulate the neutron flux distribution well with 2D Sn code, neutron flux behaviors inside the PCV had been investigated with referencing the measurement values and with observing calculated 3D neutron flux distribution. 2D calculation model had been modified repeatedly until reliable calculation result was provided. After several model modifications, the reliable 2D calculation was accomplished and important neutron transport phenomena that are necessary to simulate the neutron flux distribution well was understood. Network-parallel-computing technique was applied to radioactivity distribution calculation. Using this technique, we could calculate radioactivity at all space mesh points that were used with 2D Sn code and we obtained the radioactivity distribution. By using this distribution, we can estimate a quantity of radioactivity around RPV more accurately and optimize dismantling designs.

Solidification of Simulated Liquid Waste of Primary Loop Resin Elution Process of PWR

Primary loop resin waste is eluted by sulfuric acid in The Kansai Electic Company Mihama, Takahama and Ohi nuclear power station. Waste solution from this elution process is planned to be solidified by cement. This study bring out a range of chemical composition and crud concentration of waste solution from this elution process, and examine the properties of alumina cement solidification process and solidified material. Test for sulfate ion, borate, lithium, ammonium ion was carried out. Volume reduction ratio of over 0.5 was achieved for 5 to 25wt% of sulfate ion and <5,000ppm of borate. Lithium ion restrained the solidification delay by borate. Also, ammonium ion shows no significant effect. Based on this study, we concluded that the aluminum cement is applicable to all range of composition of waste solution from the resin elution process. This study is a part of committed work of The Kansai electric company.

Estimation and Measurement of Porosity Change in Cement Paste

Laboratory-scale experiments were performed to understand the porosity change of cement pastes. The cement pastes were prepared using commercially available Type-I ordinary Portland cement (OPC). As the cement pastes were exposed in water, the porosity of the cement pastes sharply increased; however, the slow decrease of porosity was observed as the dissolution period was extended more than 50 days. As expected, the dissolution reaction was significantly influenced by w/c raito and the ionic strength of solution. A thermodynamic model was applied to simulate the porosity change of the cement pastes. It was highly influenced by the depth of the cement pastes. There was porosity increase on the surface of the cement pastes due to dissolution of hydration products, such as portlandite, ettringite, and CSH. However, the decrease of porosity was estimated inside the cement pastes due to the precipitation of cement minerals.

Planning of Large-Scale In-Situ Gas Generation Experiment in Korean Radioactive Waste Repository

In the Korean LILW (Low- and Intermediate-Level radioactive Waste) repository at Gyeongju city, the degradation of organic wastes and the corrosion of metallic wastes and steel containers would be important processes that affect repository geochemistry, speciation and transport of radionuclides during the lifetime of a radioactive waste disposal facility. Gas is generated in association with these processes and has the potential threat to pressurize the repository, which can promote the transport of groundwater and gas, and consequently radionuclide transport. Microbial activity plays an important role in organic degradation, corrosion and gas generation through the mediation of reduction-oxidation reactions. The Korean research project on gas generation is being performed by Korea Radioactive Waste Management Corporation (hereafter referred to as “KRMC”). A full-scale in-situ experiment will form a central part of the project, where gas generation in real radioactive low-level maintenance waste from nuclear power plants will be done as an in-depth study during ten years at least. In order to examine gas generation issues from an LILW repository which is being constructed and will be completed by the end of December, 2012, two large-scale facilities for the gas generation experiment will be established, each equipped with a concrete container carrying on 16 drums of 200 L and 9 drums of 320 L of LILW from Korean nuclear power plants. Each container will be enclosed within a gas-tight and acid-proof steel tank. The experiment facility will be fully filled with ground water that provides representative geochemical conditions and microbial inoculation in the near field of repository. In the experiment, the design includes long-term monitoring and analyses for the rate and composition of gas generated, and aqueous geochemistry and microbe populations present at various locations through on-line analyzers and manual periodical sampling. A main schedule for establishing the experiment facility is as follows: Completion of the detailed design until the second quarter of the year 2010; Completion of the manufacture and on-site installation until the second quarter of the year 2011; Start of the operation and monitoring from the third quarter of the year 2011.

Verification of Source Term Analysis System for Decommissioning Wastes From a CANDU Reactor

There are now twenty commercial nuclear power reactors operating as of May 2010 in South Korea. As nuclear capacity becomes higher and installations age, the Korean government and industry have launched R&D to estimate appropriate decommissioning costs of power reactors. In this paper, MCNP/ORIGEN2 code system which is being developed as a source term evaluation tool was verified by comparing the estimated nuclide inventory from MCNP/ORIGEN2 simulation with the measured nuclide inventory from chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. Equilibrium core model of Wolsoung unit 1 was used as a neutron source to activate in-core and ex-core structural components. As a result, the estimated values from the analysis system agreed with measured data within 20% difference. Therefore, it can be concluded that MCNP/ORIGEN system could be a reliable tool to estimate source terms of decommissioning wastes from CANDU reactor, although this system assumes constant flux irradiation and snapshot equilibrium core model as a reference core.

Characterization of Radioactive Waste From Side Structural Components of a CANDU Reactor for Decommissioning Applications in Korea

Reactor core components and structural materials of nuclear power plants to be decommissioned have been irradiated by neutrons of various intensities and spectrum. This long term irradiation results in the production of large number of radioactive isotopes that serve as a source of radioactivity for thousands of years for future. Decommissioning of a nuclear reactor is a costly program comprising of dismantling, demolishing of structures and waste classification for disposal applications. The estimate of radio-nuclides and radiation levels forms the essential part of the whole decommissioning program. It can help establishing guidelines for the waste classification, dismantling and demolishing activities. ORIGEN2 code has long been in use for computing radionuclide concentrations in reactor cores and near core materials for various burn-up-decay cycles, using one-group collapsed cross sections. Since ORIGEN2 assumes a constant flux and nuclide capture cross-sections in all regions of the core, uncertainty in its results could increase as region of interest goes away from the core. This uncertainty can be removed by using a Monte Carlo Code, like MCNP, for the correct calculations of flux and capture cross-sections inside the reactor core and in far core regions. MCNP has greater capability to model the reactor problems in much realistic way that is to incorporate geometrical, compositional and spectrum information. In this paper the classification of radioactive waste from the side structural components of a CANDU reactor is presented. MCNP model of full core was established because of asymmetric structure of the reactor. Side structural components of total length 240 cm and radius 16.122 cm were modeled as twelve (12) homogenized cells of 20 cm length each along the axial direction. The neutron flux and one-group collapsed cross-sections were calculated by MCNP simulation for each cell, and then those results were applied to ORIGEN2 simulation to estimate nuclide inventory in the wastes. After retrieving the radiation level of side structural components of in- and ex-core, the radioactive wastes were classified according to the international standards of waste classification. The wastes from first and second cell of the side structural components were found to exhibit characteristics of class C and Class B wastes respectively. However, the rest of the waste was found to have activity levels as that of Class A radio-active waste. The waste is therefore suitable for land disposal in accordance with the international standards of waste classification and disposal.

Dismantling Method of Fuel Cycle Facilities Obtained by Dismantling of the JRTF

The Japan Atomic Energy Research Institute Reprocessing Test Facility (JRTF) was the first reprocessing facility which was constructed by applying only Japanese technology to establish basic technology on wet reprocessing. JRTF had been operated since 1968 to 1969 using spent fuels (uranium metal/aluminum clad, about 600kg as uranium metal and 600MWD/T) from the Japan Research Reactor No.3 (JRR-3). Reprocessing testings on PUREX process were implemented at 3 runs, so that, 200g of plutonium dioxide were extracted. After JRTF was shut down at 1970, it had been used for research and development of reprocessing since 1971. The more mature research and development of nuclear are, the more opportunity of dismantling of old nuclear facilities would be. Japan Atomic Energy Agency (JAEA) has an experience of full scale of dismantling through decommissioning of Japan Power Demonstration Reactor (JPDR)1). On the other hand, we didn’t have that of fuel cycle facility. Moreover, it is considered that dismantling methods of nuclear reactor and fuel cycle facility are different for following reason, components contaminated TRU nuclide including Pu, and components installed inside narrow cells. Dismantling methods are important factor to decide manpower and time for dismantling. So, it is indispensable to optimize dismantling method in order to minimize manpower and time for dismantling. Considering the background mentioned above, the decommissioning project of JRTF was started in 1990. The decommissioning project of JRTF is carried out phase by phase. Phase 1; Investigation for dismantling of the JRTF2)3)4). Phase 2; R&D of decommissioning technologies for dismantling of the JRTF5)6)7)8). Phase 3; Actual dismantling of the JRTF9)10). There were several components used for reprocessing and a system for liquid radwaste storage, and those were installed inside of each of several thick concrete cells. The inner surfaces of each cell were contaminated by TRU nuclides including Pu. In phase 3, components used in reprocessing and a system for liquid radwaste storage were dismantled. Moreover, opening was made in concrete walls (including ceiling) for this work. Effective practices for dismantling fuel cycle facilities were obtained through these works. On this report, effective dismantle methods obtained by actual dismantling activities in JRTF are introduced.

Implementation of Decommissioning Materials Conditional Clearance Process to the OMEGA Calculation Code

The activities performed during nuclear installation decommissioning process inevitably lead to the production of large amount of radioactive material to be managed. Significant part of materials has such low radioactivity level that allows them to be released to the environment without any restriction for further use. On the other hand, for materials with radioactivity slightly above the defined unconditional clearance level, there is a possibility to release them conditionally for a specific purpose in accordance with developed scenario assuring that radiation exposure limits for population not to be exceeded. The procedure of managing such decommissioning materials, mentioned above, could lead to recycling and reuse of more solid materials and to save the radioactive waste repository volume. In the paper an implementation of the process of conditional release to the OMEGA Code is analyzed in details; the Code is used for calculation of decommissioning parameters. The analytical approach in the material parameters assessment, firstly, assumes a definition of radiological limit conditions, based on the evaluation of possible scenarios for conditionally released materials, and their application to appropriate sorter type in existing material and radioactivity flow system. Other calculation procedures with relevant technological or economical parameters, mathematically describing e.g. final radiation monitoring or transport outside the locality, are applied to the OMEGA Code in the next step. Together with limits, new procedures creating independent material stream allow evaluation of conditional material release process during decommissioning. Model calculations evaluating various scenarios with different input parameters and considering conditional release of materials to the environment are performed to verify the implemented methodology. Output parameters and results of the model assessment are presented, discussed and concluded in the final part of the paper.