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All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Light Water Reactor Sustainability Program: Survey of Models for Concrete Degradation

Concrete is widely used in the construction of nuclear facilities because of its structural strength and its ability to shield radiation. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. As such, when life extension is considered for nuclear power plants, it is critical to have predictive tools to address concerns related to aging processes of concrete structures and the capacity of structures subjected to age-related degradation. The goal of this report is to review and document the main aging mechanisms of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

The status and prospective of environmental radiation monitoring stations in Saudi Arabia

The use of nuclear technology requires an environmental monitoring program to ensure the safety of the environment, and to protect people from the hazards of radioactive materials, and nuclear accidents. Nuclear accidents are unique, for they incur effects that surpass international frontiers, and can even have a long lasting impact on Earth. Such was the case of the Chernobyl accident in the Ukraine on April 6, 1986. For that purpose, international and national efforts come together to observe for any nuclear or radioactive accident. Many states, including Saudi Arabia which oversees the operation of the National Radiation, Environmental and Early Monitoring Stations, The Radiation Monitoring Stations(RMS’s) are currently scattered across 35 cities in the country,. These locations are evaluated based on various technological criteria such as border cities, cities of high population density, wind direction, etc. For new nuclear power plants hovering around, it is strongly recommended to increase the number of radiation monitoring stations to warn against any threat that may arise from a nuclear leak or accident and to improve the performance of the existing RMS’s. SARA (Spectroscopic Monitoring Station for air) should be implemented due to the high sensitivity to artificial radiation, automatic isotope identification, free of maintenance, and fully independent due to solar power supply (incl. battery backup) and wireless communication (GPRS)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Exploring Actinide Nanocrystal Growth towards Defining 5f Surface Chemistry (Final Technical Report)

When particles are very small, down to the nanometer length scale, they display unusual properties not typical of larger materials. Most of the atoms in these structures are at the surface of the particles, which give them different electronic properties. While these unusual properties have been studied in most of the periodic table, little is known about how the electronic properties of the actinides, such as neptunium and plutonium, behave when they become so small. This planned research will study growth pathways of actinide nanoscale particles. Interestingly, the shapes of the particles change depending on the specific actinide species, spanning from thorium through americium, despite having otherwise identical arrangements of atoms. This project will use differences in the shapes and growth pathways of different actinide particles to learn about trends in electronic properties at surfaces across the actinide row. Growth pathways will be determined using a combination of synthesis and the use of advanced X-ray characterization tools. These X-ray characterization tools can provide information about electronic properties and also local bonding characteristics. This project will also inform how to make unprecedented actinide oxide nanoparticles using very small quantities of material. This will enable safe working with radioactive materials and the study of materials that never before have been made into nanoparticles. Meanwhile, students will be trained in working with these exotic materials.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Expected Waste Streams Originating from the Decommissioning of NPPs: German Situation as Compared to International Standards - 16447

This paper presents an overview of current and future decommissioning projects of NPPs in Germany. The expected amounts of residual radioactive materials are given. Their possible utilization and disposal routes, which are stipulated by the German Radiation Protection Ordinance, are presented and compared to those resulting from the French zoning model. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Decommissioning and Demolition of a Redundant UK Research Facility at AWE Aldermaston - 12453

The redundant two-storey brick built research facility on the AWE Site at Aldermaston, UK is in the closing stages of decommissioning and demolition. The facility was used for a variety of purposes up to 1995 predominately involving the use of alpha-emitting isotopes. The two main areas of alpha-based contamination have been decommissioned with the removal of hot -boxes and fume cupboards on the ground floor and HEPA filter units and ventilation equipment on the first floor. Many of these activities were undertaken using both airline fed suits, (supplied via a free standing mobile unit), and full face respirators. Asbestos materials were located and cleared from the first floor by specialist contractor. All sections of active drain running from the building to the site active effluent disposal system were removed early in the program using established techniques with specialist monitoring equipment used to provide confidence in the data required for disposal of the decommissioning debris. In particular a dedicated High Resolution Gamma Spectrometer (radioactive materials scanning unit) was utilized to categorise waste drums and wrapped packages. The building has been decommissioned and the monitoring and sampling of the structure was completed in November 2011 - the results demonstrating that the building was clear of contamination in accordance with UK clearance and exemption requirements. The demolition plan was developed and implemented in December with site excavation of foundations and site clearance currently ongoing in preparation for final site backfill activities and project close. A number of useful lessons have been learnt during the operations and are set out at the rear of the main text. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Optimization of the Shielding of a Canister Loading Machine with the deterministic Code Attila{sup R} - Paper 61

In general shielding components are designed with very conservative assumptions and boundary conditions to cover all possibly occurring situations. This often results in an overestimated shielding design in a lot of different fields of application; the shielding of components will be heavier and more expensive than needed. In most cases an initial shielding calculation in the design process shows an unbalanced shielding effect of the shielded radiation source. Sometimes the dose rate limits are slightly exceeded in several areas. From this arises a need for an optimization of the shielding structure. The goal is to decrease the radiation level in the accessible areas to keep the dose rate limits and to reduce the amount of the used shielding material if possible. Shielding material is moved from areas with higher shielding factors to which with lower ones. The optimization process is shown in an example of a loading machine for vitrified waste in canisters. The design review with an optimization of the shielding geometry was done with the deterministic code Attila{sup R} of the company Transpire, Inc. Attila{sup R} is a powerful tool to calculate and optimize the shielding effect of handling facilities for radioactive materials. It is easy to handle with its CAD integration module and an easy setup of the calculation cases. Areas with a low shielding effect which results in an exceeding of the dose rate limits can be found quickly with the illustration due to an Iso-Dose-Rate-Surface in the 3D Model. In these areas the shielding effect has to be improved. Otherwise there is material, which has no significant influence on the shielding effect and makes the handling facilities heavier and more expensive as needed. The optimization of the shielding design of the canister loading machine in the current case results in lower dose rates outside the facility to keep the dose rate limits. The demand of shielding material could be reduced and resulted in a cost reduction on the facility itself and furthermore in the caring facilities to transport this facility (cranes, etc.). (authors)

42 ENGINEERING↗
DOE OSTI2012

Tracking and Monitoring with Dosimeter-Enabled ARG-US RFID System - 12009

Automated monitoring and tracking of materials with radio frequency identification (RFID) technology can significantly improve both the operating efficiency of radiological facilities and the application of the ALARA (as low as reasonably achievable) principle in them. One such system, called ARG-US, has been developed by Argonne National Laboratory for the U.S. Department of Energy (DOE) Packaging and Certification Program to use in managing sensitive nuclear and radioactive materials. Several ARG-US systems are in various stages of deployment and advanced testing across DOE sites. ARG-US utilizes sensors in the tags to continuously monitor the state of health of the packaging and promptly disseminates alarms to authorized users. In conjunction with global positioning system (GPS) tracking provided by TRANSCOM, the system can also monitor and track packages during transport. A compact dosimeter has been incorporated in the ARG-US tags via an onboard universal asynchronous receiver/transmitter interface. The detector has a wide measurement range for gamma radiation - from 0.1 mSv/h to 8 Sv/h. The detector is able to generate alarms for both high and low radiation and for a high cumulative dose. In a large installation, strategically located dosimeter-enabled tags can yield an accurate, real-time, 2D or 3D dose field map that can be used to enhance facility safety, security, and safeguards. This implementation can also lead to a reduced need for manned surveillance and reduced exposure of personnel to radiation, consistent with the ALARA principle at workplaces. (authors)

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DOE OSTI2014

RADTRAN 6 Technical Manual

This Technical Manual contains descriptions of the calculation models and mathematical and numerical methods used in the RADTRAN 6 computer code for transportation risk and consequence assessment. The RADTRAN 6 code combines user-supplied input data with values from an internal library of physical and radiological data to calculate the expected radiological consequences and risks associated with the transportation of radioactive material. Radiological consequences and risks are estimated with numerical models of exposure pathways, receptor populations, package behavior in accidents, and accident severity and probability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Radiological Worker II Training, Course 20301 (Live), Course 12909 (Test)

Radiological worker training is the basic building block for any additional radiological training you may receive. Upon completing radiological worker training, you will have the basic knowledge needed to work safely, using proper radiological practices, in areas where radiological hazards exist. You will also have a better understanding of the hazards and responsibilities associated with radiological work to help prevent the carelessness that can occur when working continually with or around radioactive material. This course does not qualify you for any specific radiological work. You may be required to take additional training at individual facilities to address facility- and job-specific hazards and procedures.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Treatment of Uranium Slugs at the CEA Marcoule site - 12026

The decladding units on the Marcoule nuclear site in southeast France were commissioned in the early 1960's to receive spent fuel cartridge for interim storage in pools and to prepare them for dissolution. After decladding, the uranium slugs were sent to the reprocessing plant for dissolution. The units were shut down in 1997. All the operational and process equipment was dismantled and transferred to the ANDRA waste repository between 1998 and 2001. During cleanup of the G2-G3 interim storage pool all the internal components (baskets and machinery) mounted on the civil engineering structures were cut up. In November 2007, a UNGG fuel slug in its aluminum cartridge was found at the bottom of the pool. Similarly, while iron slugs (used to control plutonium-producing reactors) were being sorted, five de-cladded uranium slugs were found between 2006 and 2008. As the processing facilities had already been dismantled, it was decided to send these fuel cartridges to a laboratory for dissolution. The project was to recover and condition the cartridge and slugs and transport them safely to the laboratory. This required the design of dedicated equipment due in particular to the high dose rate and pyro-phoricity hazard. The procedure also required safety clearance by the regulatory authorities for the operations within the facility and for the transport of radioactive materials. The cartridge was dissolved in May 2009, 18 months after it was discovered. Based on the experience acquired, the retrieval of the uranium slugs was completed in March 2011 after a year of preparation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Standards and Guidelines for the Design of Trunnions of RAM Transport Packages - 17365

Trunnion systems of packages for the transport of radioactive materials have to guarantee the safe handling of package during crane operations (lifting, tilting) and to secure package tie down to the transport vehicle, if the trunnions are used as attachment points during transport. The design of trunnions is based on IAEA Regulations SSR-6, the supporting Advisory Material SSG-26 and further appropriate technical standards and/or guidelines. For package approval procedures in Germany the guideline BAM-GGR 012 has to be applied. If a package is handled inside a german nuclear power plant, the standard KTA 3905 has to be fulfilled additionally. In this paper the requirements of KTA 3905 [4] concerning the trunnion systems as load attaching points (LAP) are discussed in connection with the recommendations in the guideline BAM-GGR 012 [3]. This guideline is prepared at BAM Federal Institute for Materials Research and Testing for analysis and assessment of bolted lid and trunnion systems of Type B(U) transport packages. The quality assurance questions concerning trunnion systems are discussed as well. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Total Integrated Dose to Spent Fuel Pool Level Instrument Probes and Refueling Floor Personnel Dose Rates Meeting NRC Order EA-12-051 for a Typical Boiling Water Reactor - Paper 22

In reviewing the events after the March 11, 2011 Fukushima accident, the Nuclear Regulatory Commission (NRC) issued Order EA-12-051, Reference 1, requiring new reliable spent fuel pool (SFP) instrumentation for pool level determination under beyond design basis event conditions. In 2013, URS prepared a calculation for a typical Boiling Water Reactor to determine the Total Integrated Dose (TID) potentially to be imparted to radiation sensitive underwater probes (one primary and one backup) planned for meeting this Order. Additionally, URS calculated the radiological dose rates to an individual standing on the refueling floor for various SFP fill levels. NEI 12-02, Reference 2, defines the levels at which SFP condition identification is necessary as the following: Level 1 - level that is adequate to support operation of the normal fuel pool cooling system. Level 1 represents the higher of the following two points: - The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or - The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis. Level 2 - level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. Level 2 is based on either of the following: - 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or - A designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components. Level 3 - level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. Level 3 corresponds nominally (i.e. +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. The URS calculation did not define these levels or determine instrumentation setpoints. However, it determined refueling floor dose rates that could be used as a tool to determine the appropriate water depths for these levels. In determining the TID for the SFP probes and the radiological dose rates to the individual, the following radioactive sources were considered: 1. Spent Nuclear Fuel (SNF) stored in the spent fuel storage racks, 2. Local Power Range Monitors (LPRMs) stored above the top of the racks on hangers on the SFP walls 3. Radioactive material laden filters stored on platforms on top of the racks, and 4. Control Rod Blades (CRBs) stored above the top of the racks on hangers on the SFP walls. The TID for the SFP probes includes both the dose from normal operation (pool full of water) for the remaining plant life (an assumed 40 years, conservatively including an allowance for lifetime extension SFP loading) and, consistent with the NEI 12-02 guidelines, Reference 2, the assumed drop in elevation of the pool water level to the top of the spent fuel racks for a 6 hour period. The acceptable TID for the radiation-sensitive probe as defined by the manufacturer was 2x10{sup 9} rad. The individual on the pool deck was assumed to be standing on the edge of the pool with an unobstructed view of its contents, as a worst case scenario. Dose rates to this individual determined from direct shine from the same 4 sources were determined for SFP water levels varying from 1 ft above the SNF racks to 21 ft above the racks in one foot increments. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Irreversible Wash Aid Additive for Cesium Mitigation. Small-Scale Demonstration and Lessons Learned

The Irreversible Wash Aid Additive process has been under development by the U.S. Environmental Protection Agency (EPA) and Argonne National Laboratory (Argonne). This process for radioactive cesium mitigation consists of a solution to wash down contaminated structures, roadways, and vehicles and a sequestering agent to bind the radionuclides from the wash water and render them environmentally immobile. The purpose of this process is to restore functionality to basic services and immediately reduce the consequences of a radiologically-contaminated urban environment. Research and development have resulted in a down-selection of technologies for integration and demonstration at the pilot-scale level as part of the Wide Area Recovery and Resiliency Program (WARRP) under the Department of Homeland Security and the Denver Urban Area Security Initiative. As part of developing the methods for performing a pilot-scale demonstration at the WARRP conference in Denver in 2012, Argonne conducted small-scale field experiments at Separmatic Systems. The main purpose of these experiments was to refine the wash water collection and separations systems and demonstrate key unit operations to help in planning for the large scale demonstration in Denver. Since the purpose of these tests was to demonstrate the operations of the system, we used no radioactive materials. After a brief set of experiments with the LAKOS unit to familiarize ourselves with its operation, two experiments were completed on two separate dates with the Separmatic systems.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Mobile platform for radiological characterization of sites under or after decommissioning - 15139

Nowadays, nuclear industry is facing a crucial need in establishing radiological characterization for the assessment and the monitoring of any remediation work. Regarding its experience in this domain, the French Commission for Atomic Energy and Alternative Energies (CEA) of Fontenay-aux-Roses, established an important feedback and developed a sound methodology for radiological characterization. A clean-up project is managed from the initial surface characterization to the final one, once volumes of contaminated soil have been removed. Samples are collected via recognized methods and data is gathered by real-time acquisition with different conveyance such as all-in-vehicles, cart, walking, helicopter or drone. Accurate radiological analyses are performed on site thanks to the mobile laboratories cluster, called SMaRT. It enables to realize long-time gamma spectroscopy, chemical preparation for alpha spectroscopy or beta liquid scintillation counting. Data is processed with geo-statistics via the software Kartotrak, 2D or 3D mapping are created with the associated uncertainty and probability map. Drilling profiles with geological data (layers composition, particles size, etc.) are studied in order to understand the behavior of the contamination in depth. Source term, contaminated volumes and sanitary impact are calculated in order to determine the optimum excavation depth. Finally, data is stored in an online site that insures the traceability of the whole process. This just-in-time production model allows a fast response and control of costs and avoids the disadvantages of radioactive material transport. This whole innovative process, developed inside the mobile platform, requires all these tools that have been used in France and continuously evolving for more than 10 years thanks to a learning organization. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Fukushima Media Involvement: Lessons Learned and Challenges - 13261

Only days after the Fukushima nuclear reactor disaster on March 11, 2011, the DOE's Pacific Northwest National Laboratory, or PNNL, found itself in a maelstrom of media attention following its announcement of the detection of minute levels of radioactive material originating from the damaged reactors 4,500 miles away. Because PNNL develops state-of-the-art ultra-sensitive radionuclide detection and monitoring systems for national security applications, and has some of the equipment operating on its Richland campus, there was little surprise when one of these sophisticated systems led PNNL to be the first to detect measurable radionuclides in the United States. On Wednesday, March 16, 2011, that system detected minuscule levels of short-lived radioactive xenon, a telltale element derived from either weapons testing or a major reactor disruption. Immediately after the detection was announced, a flurry of inquiries nearly overwhelmed staff as governments, scientific organizations, the general public, and reporters struggled to understand and estimate what impacts this disaster might have on health and environment. Over the course of about three weeks, PNNL's News and Media Relations staff and its scientists and engineers responded to more than 100 requests for information, and engaged in dozens of personal interviews with international, national, regional, and local media. While many of the interviews and resulting stories were accurate and well done, not all communication went flawlessly. In the midst of chaos and confusion, which are part of any significant crisis, hiccoughs are sure to occur. Addressed here is 'the rest of the story'. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

From the Lost Radium Files: Misadventures in the Absence of Training, Regulation, and Accountability

Purpose: Radium was the foundation of brachytherapy in the early decades of the 20th century. Despite being a most precious and perilous substance, it was mislaid with surprising frequency. This essay explores how it was lost, the efforts taken to recover it, and measures instituted to prevent mishandling. Methods and Materials: Review of contemporary literature, government publications, archives, and lay press. Results: Radium is a particularly dangerous substance because of its long half-life, its gaseous daughter (radon), and the high-energy emissions of its decay products. Despite the hazard, it was unregulated for most of the century. Any physician could obtain and administer it, and protocols for safe handling were generally lacking. Change came with appreciation of the danger, regulation, mandated training, and the institution of a culture of accountability. Unfortunately, careless management of medical radionuclides remains a global hazard. Conclusion: Responsible stewardship of radioactive material was not a high priority, for practitioners or the federal government, for much of the 20{sup th} century. As a result, large quantities of radium had gone astray, possibly subjecting the general public to continued radiation exposure. Lessons from the radium era remain relevant, as medical radionuclides are still mishandled.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Current Status of Decommissioning Plans and Chemical Decontamination of Metal Scraps in Chinshan NPP - 19304

The decommissioning of nuclear facilities has become a significantly attractive and worldwide topic due to the end of lifetime of many nuclear power plants within next coming years. In Taiwan, Chinshan nuclear power plant owned by the Taiwan Power Company will be the first candidate to be decommissioned according to the Nuclear Reactor Facilities Regulation Act. The Chinshan nuclear power plant located in New Taipei City has two units of 636 MWe boiling water reactors (BWRs) for the generation of 9 billion kWh of electricity per year. The date of cessation of operation of Unit No. 1 and Unit No. 2 will be December in 2018 and July in 2019, respectively. Based on the Regulations for the Review and Approval of Applications for a Decommissioning Permit of Nuclear Reactor Facilities, Taiwan Power Company have submitted a decommissioning plan of Chinshan nuclear power plant to the authority, Atomic Energy Council (AEC). After the review process, the decommissioning plan of Chinshan nuclear power plant has been permitted by AEC in 2017. Decontamination is generally utilized during the period of decommissioning of nuclear facility in order to minimize the radiation exposure of personnel and the cost of management of radioactive wastes. The successful decontamination also probably provides the benefits of lowering the level of radioactive wastes, facilitating potential releases, permitting the reuse of metals and facilitating the radiation principle of As Low As Reasonably Achievable (ALARA) management. Decontamination aims at the removal of radioactive materials from the contaminated surfaces by several technologies such as washing, heating, chemical reaction and mechanical process. The selection of proper decontamination technologies are depended on the requirement of the target materials and the purpose of the decontamination. Chemical decontamination has been proven to be an effective in reducing the radioactivity for metallic materials by using chemical reagents in contact with the contaminated surfaces. During the operation of Chinshan nuclear power plant, chemical decontamination was performed for Unit No.1 and No.2 in 2000 to decontaminate the Reactor Recirculation System (RRS) and Reactor Water Clean-up Systems (RWCU) by using LOMI-NP-LOMI process. Based on the previous experience of chemical decontamination, the dismantled metallic piping is provided as sample by Chinshan nuclear power plant for the evaluation of the feasibility of chemical decontamination. The provided metallic sample is composed of carbon steel with dose rate around 0.2 mSv/hr. In this report, a multistep decontamination process using strongly oxidizing solution and organic acid solution is performed to remove the contaminated oxide layer from the metal surface, and the resulting contact dose rate and decontamination factor (DF) are also shown to discuss the performance in relation to the process control factors. In this paper, the recent progress in decommissioning programs and preliminary results of the decontamination experiment are described to provide usable information regarding the future decommissioning for Chinshan nuclear power plant. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗