bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of Methodology of Groundwater Flow and Solute Transport Analysis in the Horonobe Area, Hokkaido, Japan
Authors: Maekawa Keisuke , Makino Hitoshi , Kurikami Hiroshi , Niizato Tadafumi , Inagaki Manabu , Kawamura Makoto
Published: April 2011

It is important for establishment of safety assessment techniques of geological disposal to understand groundwater flow and solute transport accurately. Therefore, we are positioning to confirm an applicability of the techniques in realistic environment as a crucial issue in R&D. We have attempted and planed some relevant studies as below: - A methodology to integrate activities from site investigations to evaluation of solute transport was examined. We have carried out groundwater flow analysis on a regional scale using geological and hydrological information from surface-based investigations at the Horonobe area, and also solute transport analysis based on the information of the trajectory analysis. - We have carried out a preliminary simulation of groundwater flow and salinity concentration distribution using information on climatic and sea-level changes, and evolution of geological structures considering the impacts of natural events and processes. Consequently, we could outline the impacts of natural events and processes on geological environment including hydrogeology, hydrochemistry and their evolutions. - We have been planning to develop and apply a methodology of groundwater flow and solute transport analysis to the shallow part, the Horonobe coastal area and around the URL. These techniques would become a basis for future site specific safety assessment in Japan.
Improvement of Radioactivity Inventory Evaluation Procedure in Preparatory Tasks for Decommissioning
Authors: Tanaka Ken-ichi , Ichige Hideaki , Tanabe Hidenori
Published: April 2011

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.
Advanced ORIENT Cycle: Progress on Fission Product Separation and Utilization
Authors: Yamagishi Isao , Ozawa Masaki , Mimura Hitoshi , Kanamura Shohei , Mizuguchi Koji
Published: April 2011

Fission reaction of U-235 and/or plutonium generates more than 40 elements and 400 nuclides in the spent fuel. Among them, 31 elements are categorized as rare metals. In a conventional fuel cycle U and Pu are reused but others are vitrified for disposal. Adv.-ORIENT (Advanced Optimization by Recycling Instructive Elements) Cycle strategy was drawn up for the minimization of radio-toxicity and volume of radioactive waste as well as the utilization of valuable elements/nuclides in the waste. The present paper describes the progress on Fission Products (FP) separation in this Cycle. Highly functional inorganic adsorbent (AMP-SG, silica gel loaded with ammonium molybdophosphate) and organic microcapsule (CE-ALG, alginate gel polymer enclosed with crown ether D18C6) were developed for separation of heat-generating Cs and Sr nuclides, respectively. The AMP-SG adsorbed more than 99% of Cs selectively from a simulated High-level Liquid Waste (HLLW). The ALG microcapsule adsorbed 0.0249 mmol/g of Sr and exhibited the order of its selectivity; Ba > Sr > Pd >> Ru > Rb > Ag. The electrodeposition is advantageous for both recovery and utilization of PGMs (Ru, Rh, Pd) and Tc because PGMs are recovered as metal on Pt electrode. Among PGMs, Pd was easily deposited on the Pt electrode. In the presence of Pd or Rh the reduction of Ru and Tc was accelerated more in hydrochloric acid media than in nitric acid. In the simulated HLLW, the redox reaction of Fe(III)/Fe(II) disturbed deposition of elements except for Pd. The deposits on Pt electrode showed higher catalytic reactivity on electrolytic hydrogen production than the original Pt electrode. The reactivity of deposits prepared from the simulated HLLW was higher than that from solution containing only PGM.
Development of Comprehensive Techniques for Coastal Site Characterisation: Part 1—Strategic Overview
Authors: Ota Kunio , Amano Kenji , Niizato Tadafumi , Alexander W. Russell , Yamanaka Yoshiaki
Published: April 2011

Any assessment of long-term repository safety will require development of a set of analyses and arguments to demonstrate the persistence of the key safety functions of the geological environment up to several hundred thousand years into the future. However, likely future global climatic and sea-level fluctuations and uplift/subsidence would result in a dramatic change in the location of the current coastline with a subsequent significant change to hydraulic and hydrochemical conditions at coastal sites. It is thus of great importance in the Japanese disposal programme to establish comprehensive techniques for coastal site characterisation. To this end, a systematic framework, which is known as a ‘Geosynthesis Data Flow Diagram’, has been formulated, which outlines a basic roadmap of the geosynthesis methodology for characterising temporal and spatial changes of various properties and processes at coastal sites, with particular focus on the palaeohydrogeology. A basic strategy for stepwise surface-based investigations has also been proposed, which incorporates the geosynthesis methodology in an effective manner. This technique has been introduced in an ongoing collaborative programme for characterising the coastal geological environment around Horonobe in northern Hokkaido, Japan, and now tested and optimised based on accumulated technical knowledge and experience during the progress of the investigations.
Natural Analogues of Cement: Overview of the Unique Systems in Jordan
Authors: Kamei G. , Alexander W. R. , Clark I. D. , Degnan P. , Elie M. , Khoury H. , Milodowski A. E. , Pitty A. F. , Salameh E. , Smellie J. A. T.
Published: April 2011

In many radioactive waste repository designs, cement-based materials are expected to dominate the repository and models of cement evolution predict that leaching of the cementitious material in the repository by groundwater will produce an initial stage of hyperalkaline (pH∼13.3) leachates, dominated by alkali hydroxides, followed by a longer period of portlandite and C-S-H (CaO-SiO2-H2O) buffered (pH∼12.5) leachates. It has also been predicted that, as the hyperalkaline porewater leaches out of the near-field, significant interaction with the repository host rock and bentonite buffer and backfill may occur. This could possibly lead to deterioration of those features for which the host rock formation and bentonite were originally chosen (e.g. low groundwater flux, high radionuclide retardation capacity etc). The precise implications of cement leachate/repository host rock interaction has been studied in the laboratory and in underground research laboratories (URLs) and this work has been supported by study of natural cements in Jordan. These natural cements have been produced by the combustion of organic-rich clay biomicrites and are very close analogues of industrial cement. Following interaction with groundwaters, natural hyperalkaline leachates are produced and these move out of the cement into the surrounding host rock, subsequently interacting with and altering it.
Development of a Quality Management System (QMS) for Borehole Investigations: Part 2—Evaluation of Applicability of QMS Methodology for the Hydrochemical Dataset
Authors: Kunimaru Takanori , Ota Kunio , Amano Kenji , Alexander W. Russell
Published: April 2011

An appropriate QMS, which is among the first tools required for repository site characterisation, will save on effort by reducing errors and the requirement to resample and reanalyse–but this can only be guaranteed by continuously assessing if the system is truly fit-for-purpose and amending it as necessary based on the practical experience of the end-users on-site. A QA audit of hydrochemical datasets for boreholes HDB-1–11 from Horonobe URL project by JAEA has been carried out by the application of a formal QA analysis which is based on the methodology previously employed for groundwaters during the recent site characterisation programme in Sweden. This methodology has been successfully applied to the groundwaters of the fractured crystalline rocks of the Fennoscandian Shield and has now been adapted and applied to some of the ground- and porewaters of the Horonobe URL area. This paper will present this system in the context of the Japanese national programme and elucidate improvements made during hands-on application of the borehole investigation QMS. Further improvements foreseen for the future will also be discussed with a view of removing inter-operator variability as much as is possible. Only then can confidence be placed in URL project or repository site hydrochemical datasets.
Technical Know-How for Modeling of a Geological Environment: Part 2—Geological Modeling
Authors: Matsuoka Toshiyuki , Amano Kenji , Osawa Hideaki , Semba Takeshi
Published: April 2011

It is important for site characterization projects to manage the decision-making process with transparency and traceability and to transfer the technical know-how developed and accumulated during the research and development to the implementing phase as well as to future generations. The modeling of a geological environment supports efforts to clarify the degree of understanding regarding that geological environment, including uncertainty. Evaluation of the impact of uncertainties in a geological environment model is important to identify and prioritize key issues for further investigations. Therefore, a plan for site characterization should be made based on the results of the modeling. The aim of this study is to support the planning of initial surface-based site characterization based on the technical know-how accumulated from the Mizunami Underground Research Laboratory Project and the Horonobe Underground Research Laboratory Project. These projects are broad scientific studies of the deep geological environment that are a basis for research and development for the geological disposal of high-level radioactive wastes. In this study, the work-flow followed in developing the geological model, one of the geological environment models, and the related technical know-how acquired from literature data have been summarized.
Uranium Refining and Conversion Plant Decommissioning Project
Authors: Zaima Naoki , Morimoto Yasuyuki , Sugitsue Noritake , Kado Kazumi
Published: April 2011

The uranium refining and conversion plant (URCP) at Ningyo-toge was constructed in 1981 for the purpose of demonstrating on refining and conversion process from yellow cake (or uranium trioxide) to uranium hexafluoride by way of uranium tetrafluoride. For 20 years, 385 tons of natural uranium hexafluoride and 336 tons of reprocessed uranium hexafluoride (approximately) was produced. There are two different type of refining processes in the URCP. One is the wet process by convertig the natural uranium and the other is the dry conversion process for the reprocessed uranium. The dismantling of the dry process facilities began in March, 2008. It was found the large amount of uranium residuals such as wet slurry and powder uranium inside the vessels and pipes. Therefore, we have to take care of the spread of the contamination during dismantling works. The basic strategy concerning plant dismantling were the optimization of the total labor costs and the minimization of the radioactive wastes generated. The dismantling procedure is shown below; i) measuring doserate by using high sensitivity surveymeters, and nuclide identification by using gamma ray spectrometry, ii) estimating uranium mass inventory, iii) planning work force distributions with radiological survey staffs, iv) deciding dismantling methods concretely, v) decontaminating schematically if required, vi) collecting detailed data of working conditions, vi) measuring and classifying contaminated materials, vii) managements of radioactive waste drum and non-contaminated equipment, viii) control for personal exposures. Almost all equipment will be decontaminated except building decontamination it by around 2013FY. In addition, the secondary wastes were also yielded. Few thousands man-days were necessary for this project. The measurement data have not showed the high environmental radiation doserate, generally less than 0.3μSv/h. However, by the trace of the reprocessed uranium, the trans-uranium nuclides such as uranium-232 progenies, Th-228 and Tl-208 were observed. The accumulation of the nuclides which emit high energy gamma rays such as Tl-208 caused radiation exposure. As for the waste disposal, the determination of uranium content must be necessary. We have been now developing the uranium measuring systems with better accuracy. The further tasks imposed by our experiences are summarized the followings; i) minimization and reduction of radioactive wastes, ii) decontamination for the buildings and utilities, iii) wastes disposal. We have to work hard toward the final decommissioning.
Evaluation of the Long-Term Evolution of the Groundwater System in the Mizunami Area, Japan
Authors: Mizuno Takashi , Milodowski Antoni E. , Iwatsuki Teruki
Published: April 2011

This study aimed to develop a methodology for assessing the evolution of the long-term groundwater system, using fracture-filling calcite. Fracture-filling calcite mineralization in deep (to ca. 1000 m) granitic rocks in Mizunami area, Japan, was studied. Four generations (I to IV) of calcite precipitation can be differentiated based on their paragenetic relationships, morphological and isotopic characteristics. Carbon and oxygen isotopic ratios suggest that the Calcite I is of hydrothermal origin. On the other hand, Calcite II, IV and III were precipitated from freshwater and marine water, respectively. The Mizunami Group strata (Tertiary), which overly the basement Toki granite (Cretaceous), were initially deposited in a lacustrine environment but later became marine. Lacustrine conditions were re-established during the deposition of the upper Seto Group (Quaternary). It is suggested that both of deposition of the marine upper part of the Mizunami Group and the precipitation of Calcite III were possibly related to the same transgression event. This was followed by the precipitation of Calcite IV during subsequent fresh water flushing of the earlier marine groundwater. In summary, integrated morphological, mineralogical, microchemical and isotopic analysis of multilayered calcite fracture mineralization provides valuable information to evaluate the long-term evolution of groundwater system.
Reuse of Conditionally Released Radioactive Materials From NPP Decommissioning Applied in Motorway Bridges Construction
Authors: Panik Michal , Hrncir Tomas , Necas Vladimir
Published: April 2011

During the operation and especially during decommissioning of nuclear installation is produced considerable amount of solid materials (metals, non-metals, building structures) that can fix radioactivity in forms of contamination or activation. The materials present radioactive waste, part of radioactive waste may just slightly exceed limits for unconditional release of materials into the environment. On the other side, there is possible, after proving of defined safety limits fulfillment, to conditionally release radioactive waste for special purpose. In opposite case it would be inevitable to dispose radioactive waste in radioactive waste repository. Approaches of different countries to release of materials vary and the extent of this issue processing is related to each approach. Requirements set down in Slovak Republic legislation are given in the paper. Before the conditional release of materials there must be done consistent analysis of the materials impact on the inhabitants and the environment in short and long time period. The analysis comprises the evaluation of considered scenarios of specific utilization of conditionally released materials. This analysis necessarily precedes the realization of utilization. Scenarios describing utilization of radioactive waste carbon steel in the motorway bridge building process is stated in the paper. Radioactive steel can be utilized in many parts of the bridge. In the paper it is described its use as reinforcement of piles. Short time period external irradiation of workers and inhabitants is taken into account. Critical group (i.e. the group that gets the highest accumulated dose) of workers or inhabitants is chosen. Specific mass activity of released radioactive waste carbon steel is related to individual effective dose taken by critical group. Following legislation rules, annual effective dose taken by critical group must not overstep the limit of 10 μSv/year. The determination of value of the specific mass activity is the target of scenarios evaluating. Evaluation of model scenarios can be realized with the appropriate calculation tool. In the paper VISIPLAN 3D ALARA planning tool was chosen.
Development of New Ultrafiltration Techniques Maintaining In-Situ Hydrochemical Conditions for Colloidal Study
Authors: Aosai Daisuke , Yamamoto Yuhei , Mizuno Takashi
Published: April 2011

Chemical state of elements in groundwater is one of the most important information for understanding behavior of elements in underground environment. Chemical state of elements controlled mainly by groundwater physico-chemical parameters. Because the change of physico-chemical parameters of groundwater, due to pressure release and oxidation during sampling, causes changes in chemical state of elements, systematic methodologies for understanding in situ chemical state is required. In this study, in order to understand chemical state of elements in groundwater, an ultrafiltration instrument for maintaining in-situ pressure and anaerobic conditions was developed. The instrument developed in this study for ultrafiltration made of passivated Stainless Used Steel (SUS) materials, was designed to keep groundwater samples maintaining in-situ pressure/anaerobic conditions. Ultrafiltration of groundwater was conducted at a borehole drilled from the 200 mbGL (meters below ground level) Sub-stage at a depth of 200 m at the Mizunami Underground Research Laboratory. Chemical analyses of groundwater were also conducted using samples filtered under both pressurized/anaerobic and atmospheric conditions and passivated SUS materials with different elapsed times after passivation. The results indicate that our ultrafiltration method is suitable for collection of filtered groundwater and passivation is an essential treatment before ultrafiltration.
Methods of Selected Input Calculation Data Verification and Their Influence on Decommissioning Cost in the Omega Code
Authors: Ondra Frantisˇek , Danisˇka Vladimi´r , Reha´k Ivan , Necˇas Vladimi´r , Schultz Oto
Published: April 2011

The aim of this contribution is development of methodology for verification of selected input calculation data (performance unit parameters, work group structure, and duration of time-dependent activities) of the OMEGA Code in the individual PSL (Proposed Standardised List) structure parts.
Territorial Integration of the Geological Repository in France
Authors: Farin S. , Ouzounian G. , Miguez R. , Tison J.-L.
Published: April 2011

In France, a framework has been drawn up by the National Assembly and implemented by the government, in order to get the best relationship between Andra, among others, and the stakeholders and the inhabitants of the towns and countries where disposal facilities or projects are or could be established. The main threads of the two Acts passed in 2006, being relevant to the relationship with inhabitants are the information exchange and the local economic development. Dealing with the information exchange and diffusion: • The Local Information Committee (CLI), for each nuclear facility, has been reinforced and a specific, Local Information and Oversight Committee (for the Underground Laboratory in Meuse-Haute-Marne) has been renewed. The CLI was in charge of a general assignment to inform and consult on nuclear safety, radioprotection and environmental topics. Now, since 2006, the nuclear facility’s CLI and the CLIS are able to order study reports, measures and analyses to experts freely selected. • Creation of the High Committee for Transparency and Information on Nuclear Safety (HCTISN). This new authority aims to inform, consult and debate about the risks relevant to nuclear activities and their impacts on people’s health, environment and nuclear safety. Andra contributes to the functioning of CLI in disposal facilities at Manche and Aube Departments, and CLIS of the underground Laboratory at Meuse and Haute-Marne departments. This paper will present these contributions and how Andra’s action helps to reach the goals of information and exchange with the people around its facilities. Concerning the local economic development, there are specific organizations or schemes, depending on the facility: • Local taxes contributions based on the disposal facilities activities as is usual in France. • A High Level Committee (CHN) and two public interest groups (GIP) in Meuse and Haute-Marne departments have been set up since 1991 and 2005. Andra is represented in these three institutions, but they are not funded at all by Andra. This paper will show the Andra’s involvement in the local economic and territorial developments. Within this general framework Andra has developed information and exchanges actions with the stakeholders and the inhabitants around its facilities. Examples of these actions will be presented also.
Study on the Estimation Error Caused by Using One-Dimensional Model for the Evaluation of Dipole Tracer Test
Authors: Ijiri Yuji , Naemura Yumi , Amano Kenji , Maekawa Keisuke , Sawada Atsushi , Ota Kunio , Kunimaru Takanori
Published: April 2011

In-situ tracer tests are a valuable approach to obtain parameters for a performance assessment of nuclear waste repository. A one-dimensional model is simple and is commonly used to identify radionuclide transport parameters by fitting breakthrough curves simulated using the model to those obtained from tracer tests. However, this method can increase uncertainty and introduce errors in the estimated parameters. In particular, such uncertainties and errors will be significant when evaluating parameters for tests conducted in a dipole (two-dimensional) flow field between injection and withdrawal wells. This paper describes a numerical analysis investigation into the effects of various experimental conditions on parameters estimated using a one-dimensional model for cases involving tracer tests in a two-dimensional fracture plane. Results show that longitudinal dispersivity tends to be overestimated by the one-dimensional model analysis. This overestimation is the result of several factors: smaller pumping rate, larger dipole ratio, stronger heterogeneity of the fracture hydraulic conductivity, and greater orthogonally-oriented background groundwater flow. Such information will help us to better plan and design tracer tests at underground research laboratories located in both Mizunami in central Japan and Horonobe in northern Japan. Understanding appropriate experimental conditions will help decrease the uncertainty in the results of tracer tests.
Separation and Recovery of Sodium Nitrate From Low-Level Radioactive Liquid Waste by Electrodialysis
Authors: Meguro Yoshihiro , Kato Atsushi , Watanabe Yoko , Takahashi Kuniaki
Published: April 2011

An advanced method, in which electrodialysis separation of sodium nitrate and decomposition of nitrate ion are combined, has been developed to remove nitrate ion from low-level radioactive liquid wastes including nitrate salts of high concentration. In the electrodialysis separation, the sodium nitrate was recovered as nitric acid and sodium hydroxide. When they are reused, it is necessary to reduce the quantity of impurities getting mixed with them from the waste fluid as much as possible. In this study, therefore, a cation exchange membrane with permselectivity for sodium ion and an anion exchange membrane with permselectivity for monovalent anion were employed. Using these membranes sodium and nitrate ions were effectively removed form a sodium nitrate solution of high concentration. And also it was confirmed that sodium ion was successfully separated from cesium and strontium ions and that nitrate ion was separated from sulfate and phosphate ions.
AREVA NP Decontamination Concept for Decommissioning: A Comprehensive Approach Based on Over 30 Years Experience
Authors: Stiepani Christoph
Published: April 2011

Decontamination prior to Decommissioning and Dismantlement is imperative. Not only does it provide for minimization of personnel dose exposure but also maximization of the material volume available for free release. Since easier dismantling techniques in lower dose areas can be applied, the licensing process is facilitated and the scheduling and budgeting effort is more reliable. The most internationally accepted approach for Decontamination prior to Decommissioning projects is the Full System Decontamination (FSD). FSD is defined as the chemical decontamination of the primary cooling circuit, in conjunction with the main auxiliary systems. AREVA NP has long-term experience with Full System Decontamination for return to service of operating nuclear power plants as well as for decommissioning after shutdown. Since 1976, AREVA NP has performed over 500 decontamination applications and, from 1986, Decontaminations prior to Decommissioning projects which comprise virtually all NPP designs and plant conditions were performed: • NPP designs: HPWR, PWR, and BWR by AREVA, Westinghouse, ABB and GE; • Decontaminations performed shortly after final shutdown or several years later, and even after re-opening Safe Enclosure; • High Alpha inventory and or low gamma/alpha ratio; • Main Coolant chemistry (e.g., with and without Zn injection during operation). Fifteen Decontaminations prior to Decommissioning Projects have been performed successfully to date and the sixteenth FSD is now in the detailed engineering phase and is scheduled to commence late 2010. AREVA NP has developed a fully comprehensive approach for decontamination based on the CORD® (Chemical Oxidation Reduction Decontamination) Family, applied using the in-house designed decontamination equipment AMDA™ (Automatic Modular Decontamination Appliance). Based on the vast experience of AREVA NP in the field of decontamination, the Decontamination Concept for Decommissioning was developed. This concept ensures that the decontamination is tailored to the given boundaries and desired goals to ensure the best results with the lowest waste generation. This includes lower source term by oxide film, thus corrosion product, removal; controlled base metal attack to remove embedded activity; increased gamma/alpha ratios; and alpha contamination removal. This paper will describe the AREVA NP Decontamination Concept for Decommissioning (DCD) and present highlights of previous FSDs performed prior to decommissioning using the CORD/AMDA technology.
Legacy Management: Turning Liabilities Into Assets
Authors: Legare Joseph A. , Olson Eric
Published: April 2011

The Legacy Management (LM) Program has responsibility for management of over 85 post-closure sites across the United States. The program was formed through a consolidation of AEC/DOE sites being managed under separate programs and with diverse geography, regulatory bases, residual contaminants, and operating histories. Through development and implementation of a nation-wide program to ensure public safety, remedy performance, compliance, records management and ongoing stakeholder communication, the program has become efficient at meeting post-closure responsibilities and effective at proactively turning these liabilities into assets.
Determination of Environmental Uranium Concentration by Utilizing Gamma-Ray Emission From the Progeny Radionuclides
Authors: Tanaka Tadao , Shimada Taro , Ito Takeshi , Sukegawa Takenori
Published: April 2011

Nuclear facility sites such as an enrichment plant and fabrication plant are allowed to be released from the regulatory control of nuclear safety after the plants are decommissioned. It is necessary to confirm that the site has been decontaminated successfully, prior to be released. A determination method for U-238 concentration of background level in environment and for probate of vast site areas was proposed, in which the gamma-ray emission from the progeny radionuclides of U-238, such as Th-234, Pa-234m, Ra-226, are measured by in-situ gamma-ray spectrometry with a portable germanium semiconductor detector (portable Ge detector). Validity of the determination method of U-238 concentration from the progeny radionuclides was examined by the comparison between the U-238 concentration determined by the in-situ measurement with the portable Ge detector and that directly measured by ICP-MS. The U-238 concentration by the in-situ measurement was determined from peak counting rate at 186 keV of the gamma-ray emission corresponding to Ra-226. The determined U-238 concentration was in the order of 0.01 Bq/g in radioactive concentration, and was in comparable level with the concentrations decided by the ICP-MS. The proposed method utilizing gamma-ray estimation from the progeny radionuclides may be available for the U-238 concentration determination in vast land areas.
Development of the Quality Management System for Borehole Investigations: Part 1—Quality Assurance and Quality Control Methodology for Hydraulic Packer Testing
Authors: Takeuchi Shinji , Kunimaru Takanori , Ota Kunio , Frieg Bernd
Published: April 2011

A quality assurance and quality control (QA/QC) system for the hydraulic packer tests has been established based on the surface-based investigations at JAEA’s underground research laboratories in Mizunami and Horonobe. The established QA/QC system covers field investigations (data acquisition) and data analysis. For the field investigations, the adopted procedure is selection of a test section based on a detail fluid logging and checking with tally list, followed by inspection of test tools such as pressure transducers and shut-in valves, etc., test method selection using a “sequential hydraulic test” for deciding appropriate method, and finally data quality confirmation by pressure changes and derivatives on a log-log plots during testing. Test event logs should also be described during testing for traceability. For the test data analysis, a quick analysis for rough estimation of hydraulic parameters, and a detailed analysis using type curve and/or numerical analyses are conducted stepwise. The established QA/QC system has been applied to the recent borehole investigations and its efficiency has been confirmed.
Removal of Fluorine and Boron From Groundwater Using Radiation-Induced Graft Polymerization Adsorbent at Mizunami Underground Research Laboratory
Authors: Iyatomi Yosuke , Hoshina Hiroyuki , Seko Noriaki , Kasai Noboru , Ueki Yuji , Tamada Masao
Published: April 2011

High fluorine and boron contents in groundwater are commonly reduced using coagulation and ion-exchange treatments. As an alternative, we tested the efficiency of fluorine and boron removal from groundwater using radiation-induced graft polymerization adsorbent. The durability of the adsorbent was also determined by varying groundwater flow-through rates and repetitive use of the adsorbent. The results indicated that it was possible for the adsorbent to remove more than 95% of boron and fluorine from the groundwater, and that the performance of the adsorbent for boron removal was better than commonly used ion-exchange resin. The adsorbent used several times was able to remove boron, indicating that the adsorbent can be used for efficient boron removal.
Study on Evaluation Models of Management Data for Decommissioning of FUGEN
Authors: Shibahara Yuji , Izumi Masanori , Nanko Takashi , Tachibana Mitsuo , Ishigami Tsutomu
Published: April 2011

In the Fugen nuclear power plant (FUGEN), the dismantling of equipments in the turbine building has started in 2008, and the dismantling of equipments around the reactor is scheduled around in 2015. To evaluate the management data on this dismantling of equipments around reactor appropriately, it is very important to study whether the conventional evaluation models have the applicability for FUGEN or not. Thus, the management data on the dismantling of equipments in 3rd/4th feedwater heater room conducted in 2008 was calculated with the conventional evaluation models. The conventional evaluation models were made by data obtained from Japan Power Demonstration Reactor (JPDR) decommissioning program. It was found that there were large differences between the calculated values and the actual data. For finding the cause in the difference between them, the dismantling of equipments in 3rd/4th feedwater heater room was divided into three processes: i) the preparation process, ii) the dismantling process, and iii) the clean-up process. In the both process of preparation and clean-up, the calculated values were smaller than the actual data. These were mostly caused by the plant scale difference between JPDR and FUGEN, because the conventional evaluation models were built by analyzing the actual data on the decommissioning of JPDR which is smaller than FUGEN. The primary expression depending on the area of working space was built as new evaluation models for the preparation and clean-up processes. In the dismantling process, on the other hand, it was found that there were characteristic differences in the dismantling of feedwater heater as follows: 1) the calculated values were significantly larger than the actual data, 2) the actual data for the dismantling of 3rd feedwater heater was larger than that of 4th one, though these equipments were almost same weight. It was found that these were brought by the difference in the descriptions of dismantling of feedwater heaters, and the new evaluation models reflecting the descriptions of dismantling were built for the appropriate evaluation of the management data. The calculated values with the new evaluation models for each process showed the good agreement with the actual data. In this report, study on evaluation models of management data for dismantling of equipments in the feedwater heater room will be described.
Preparation of Reference Materials on Radiochemical Analysis for Low-Level Radioactive Waste Generated From Japan Atomic Energy Agency
Authors: Ishimori Ken-ichiro , Kameo Yutaka , Nakashima Mikio , Takahashi Kuniaki
Published: April 2011

In order to prepare simulated reference materials of solidified products using a laboratory-scale electric furnace instead of a plasma heating device, melting tests of miscellaneous simulated solid wastes were conducted. Over 90% of Cs was remained in the solidified product by keeping the basicity (CaO wt%/SiO2 wt%) to be 0.05. Under the optimum melting conditions, reference materials containing α-ray emitting nuclides (237Np, 241Am and 244Cm) or γ-ray emitting nuclides (60Co, 137Cs and 152Eu) were prepared. For reference materials containing β-ray emitting nuclides such as 14C and 36Cl, a preparation method with nuclear reaction 14N(n, p)14C or 35Cl(n, γ)36Cl by thermal neutron irradiation was attempted. The radioactivity concentrations of the reference materials were evaluated from radiochemical analysis. It was confirmed that the reference materials could be successfully prepared on the present preparation methods.
The Treatment of Hexavalent Chromium in Waste Liquid From Fugen Decommissioning
Authors: Ishizuka Nobuo , Sato Yuji , Fujiwara Wataru , Yahiro Yuki , Yamamoto Seiji , Negishi Koji , Fukushima Tadashi , Sakai Hitoshi , Yoshida Norimasa
Published: April 2011

Authors have developed the new treatment technique for hexavalent chromium (Cr6+) in the liquid waste. Cr6+ can be reduced to trivalent chromium (Cr3+) by hydrogen peroxide (H2O2) under acidic condition with small amount of secondary waste. In 2010, this technique was applied to the cooling water for iron-water shield surrounding the core region in Fugen Nuclear Power Station of Japan Atomic Energy Agency (JAEA). The concentration of Cr6+ and total chromium (Cr) were reduced to less than 0.5ppm and 2.0ppm in liquid waste, respectively. These values satisfy the water pollution prevention law.
Regulatory Research for Geological Disposal of High-Level Radioactive Waste in Japan
Authors: Nakayama Shinichi , Watanabe Yoshio , Kato Masami
Published: April 2011

The Nuclear and Industrial Safety Agency of the Ministry of Economy, Trade and Industry (NISA) has renewed its regulatory role and its need for regulatory research on radioactive waste management, with recent circumstances of radioactive waste management in Japan being taken into consideration. In response, a technical supporting organization, the Japan Nuclear Energy Safety Organization (JNES), in 2009 released the five-year research plan “Regulatory Research Plan on Radioactive Waste Management 2010–2014”, in cooperation with the research institutes of the Japan Atomic Energy Agency (JAEA) and the National Institute of Advanced Industrial Science and Technology (AIST). The geological disposal research plan and the future research activities are outlined in this paper. JNES launched safety studies on geological disposal in 2003, the year it was established. JAEA and AIST joined as regulatory support research institutes in 2005. In October 2007, all three parties signed an agreement of cooperative study on geological disposal, which facilitated joint studies and exchanges of staff, data, and results. One of the ongoing joint studies has focused on regional-scale hydrogeological modeling using JAEA’s Horonobe Underground Research Laboratory.
Implementation of Decommissioning Materials Conditional Clearance Process to the OMEGA Calculation Code
Authors: Zachar Matej , Danisˇka Vladimi´r , Necˇas Vladimi´r
Published: April 2011

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.
Methodology of Personnel Exposure Calculation and Optimisation Within the Decommissioning Planning Code OMEGA
Authors: Vasˇko Marek , Danisˇka Vladimi´r , Reha´k Ivan , Necˇas Vladimi´r
Published: April 2011

Calculation of personnel exposure is a one of the main parameters being evaluated within the pre-decommissioning plans together with other decommissioning drivers such as costs, manpower, amounts of RAW and conventional waste and amount of discharged gaseous and liquid effluents. Alongside with manpower, the exposure is an indicator of the decommissioning process for need of staff, and quantifies impact of decommissioning on personnel from the radio hygienic point of view. At the same time it indicates suitability of individual work procedures use for decommissioning activities. For this reason it is important to estimate as precise as possible demands on personnel exposure even during preparatory decommissioning phase to quantify impact of decommissioning on personnel and eventually optimize the decommissioning process, if needed. The most appropriate way of staff exposure estimation during decommissioning preparatory phases is its calculation based on radiological and physical characteristics of equipment to be decommissioned and also quantitative and qualitative characterisation of typical decommissioning activities. On one hand, the methodology of exposure calculation should allow as much as possible realistic description and algorithmisation of exposure ways during decommissioning activities. On the other hand the calculation have to be systematic, well-arranged and clearly definable by appropriate mathematic relations. Calculation can be made by various approaches using more or less sophisticated software solutions from classic MS Excel sheets up to the complex calculation codes. In this paper, a methodology used for personnel exposure calculation and optimization implemented within the complex computer code OMEGA [1] developed at DECOM, a.s. is described.
Program Change Management During Nuclear Power Plant Decommissioning
Authors: Bushart Sean , Kim Karen , Naughton Michael
Published: April 2011

Decommissioning a nuclear power plant is a complex project. The project involves the coordination of several different departments and the management of changing plant conditions, programs, and regulations. As certain project Milestones are met, the evolution of such plant programs and regulations can help optimize project execution and cost. This paper will provide information about these Milestones and the plant departments and programs that change throughout a decommissioning project. The initial challenge in the decommissioning of a nuclear plant is the development of a definitive plan for such a complex project. EPRI has published several reports related to decommissioning planning. These earlier reports provided general guidance in formulating a Decommissioning Plan. This Change Management paper will draw from the experience gained in the last decade in decommissioning of nuclear plants. The paper discusses decommissioning in terms of a sequence of major Milestones. The plant programs, associated plans and actions, and staffing are discussed based upon experiences from the following power reactor facilities: Maine Yankee Atomic Power Plant, Yankee Nuclear Power Station, and the Haddam Neck Plant. Significant lessons learned from other sites are also discussed as appropriate. Planning is a crucial ingredient of successful decommissioning projects. The development of a definitive Decommissioning Plan can result in considerable project savings. The decommissioning plants in the U.S. have planned and executed their projects using different strategies based on their unique plant circumstances. However, experience has shown that similar project milestones and actions applied through all of these projects. This allows each plant to learn from the experiences of the preceding projects. As the plant transitions from an operating plant through decommissioning, the reduction and termination of defunct programs and regulations can help optimize all facets of decommissioning. This information, learned through trial in previous plants, can be incorporated into the decommissioning plan of future projects so that the benefits of optimization can be realized from the beginning of the projects. This process of the collection of information and lessons learned from plant experiences is an important function of the EPRI Decommissioning Program.
EPRI Nuclear Power Plant Decommissioning Technology Program
Authors: Kim Karen S. , Bushart Sean P. , Naughton Michael , McGrath Richard
Published: April 2011

The Electric Power Research Institute (EPRI) is a non-profit research organization that supports the energy industry. The Nuclear Power Plant Decommissioning Technology Program conducts research and develops technology for the safe and efficient decommissioning of nuclear power plants.
CEA’s Radioactive Waste and Unused Fuel Inventory: Marcoule Site Example
Authors: Butez Marc , Fulleringer Daniel
Published: April 2011

In the field of clean up and decommissioning of nuclear facilities the knowledge of the radioactive waste and unused fuel inventories is crucial. For CEA (French Atomic Energy and alternative energy Commission), this knowledge is a key issue: in the short term for the flows stream lining as well as in the middle and long terms for investment strategy in the nuclear support facilities.