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At least 1,315 records · Page 73
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity of the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Use of scaled BWR lower plenum boron mixing tests to qualify the boron transport model used in TRACG

In 2001 GEH applied best estimate methods combined with a statistical methodology to determine upper bound limits for key licensing parameters for anticipated operation occurrence (AOO) transient and anticipated transients without scram (ATWS) overpressure analyses for operating Boiling Water Reactors (BWRs). The methodology was subsequently extended for ESBWR AOO, ATWS, loss of coolant, and stability analyses. GEH is extending the methodology to long-term ATWS analyses for the operating BWRs. A long-term ATWS scenario uses injection of borated water to achieve reactor shutdown. Predicting the mixing and transport of boron is important for calculating the impact on the key licensing parameters. For the many operating BWRs where the denser boron solution is injected into the lower plenum, stratification may occur, delaying boron transport to the core region. CFD modeling can be used to model the stratification and mixing of the boron solution, but such calculations are extremely computer intensive and not cost effective; therefore, a more-empirical approach supported by a theoretical scaling of the dominant phenomena and backed by test data and benchmark calculations is used. The paper presents the TRACG lower plenum boron transport model qualification effort. The scaling basis used to implement the TRACG boron transport model for BWR applications is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

CFD analyses of natural circulation in the air-cooled reactor cavity cooling system

The Natural Convection Shutdown Heat Removal Test Facility (NSTF) is currently being built at Argonne National Laboratory, to evaluate the feasibility of the passive Reactor Cavity Cooling System (RCCS) for Next Generation Nuclear Plant (NGNP). CFD simulations have been applied to evaluate the NSTF and NGNP RCCS designs. However, previous simulations found that convergence was very difficult to achieve in simulating the complex natural circulation. To resolve the convergence issue and increase the confidence of the CFD simulation results, additional CFD simulations were conducted using a more detailed mesh and a different solution scheme. It is found that, with the use of coupled flow and coupled energy models, the convergence can be greatly improved. Furthermore, the effects of convection in the cavity and the effects of the uncertainty in solid surface emissivity are also investigated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

UP2 400 High Activity Oxide Legacy Waste Retrieval Project Scope and Progress-13048

The High Activity Oxide facility (HAO) reprocessed sheared and dissolved 4500 metric tons of light water reactor fuel the fuel of the emerging light water reactor spent fuel between 1976 and 1998. Over the period, approximately 2200 tons of process waste, composed primarily of sheared hulls, was produced and stored in a vast silo in the first place, and in canisters stored in pools in subsequent years. Upon shutdown of the facility, AREVA D and D Division in La Hague launched a thorough investigation and characterization of the silos and pools content, which then served as input data for the definition of a legacy waste retrieval and reconditioning program. Basic design was conducted between 2005 and 2007, and was followed by an optimization phase which lead to the definition of a final scenario and budget, 12% under the initial estimates. The scenario planned for the construction of a retrieval and reconditioning cell to be built on top of the storage silo. The retrieved waste would then be rinsed and sorted, so that hulls could subsequently be sent to La Hague high activity compacting facility, while resins and sludge would be cemented within the retrieval cell. Detailed design was conducted successfully from 2008 until 2011, while a thorough research and development program was conducted in order to qualify each stage of the retrieval and reconditioning process, and assist in the elaboration of the final waste package specification. This R and D program was defined and conducted as a response and mitigation of the major project risks identified during the basic design process. Procurement and site preparatory works were then launched in 2011. By the end of 2012, R and D is nearly completed, the retrieval and reconditioning process have been secured, the final waste package specification is being completed, the first equipment for the retrieval cell is being delivered on site, while preparation works are allowing to free up space above and around the silo, to allow for construction which is scheduled to being during the first semester of 2013. The elaboration of the final waste package is still undergoing and expected to be completed by then end of 2013, following some final elements of R and D required to demonstrate the full compatibility of the package with deep geological repository. The HAO legacy waste retrieval project is so far the largest such project entering operational phase on the site of La Hague. It is on schedule, under budget, and in conformity with the delivery requirements set by the French Safety Authority, as well as other stakeholders. This project paves the way for the successful completion of AREVA La Hague other legacy waste retrieval projects, which are currently being drafted or already in active R and D phase. (authors)

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DOE OSTI2013

Sharing Experiences within AREVA D and D Project Portfolio: Four Illustrations - 13049

Over the past ten years, AREVA has performed D and D operations on a wide range of nuclear sites, such as Marcoule and La Hague recycling plants, to Cadarache MOX fuel fabrication plant or Veurey and Annecy metallic Uranium machining plants. Each site is different from the other but some lessons can be shared through this D and D portfolio. In that respect, knowledge management is one of AREVA D and D Technical Department main missions. Four illustrations demonstrate the interest of knowledge share. Waste management is one of the key activities in D and D; It requires a specific characterization methodology, adapted logistics, and optimized waste channels, all of which have been developed over the years by AREVA teams on the site of Marcoule while they are rather new to La Hague, whose main activity remains fuel reprocessing despite the launch of UP2 400 D and D program. The transfer of know how has thus been organized over the past two years. Plasma cutting has been used extensively in Marcoule for years, while prohibited on the site of La Hague following questions raised about the risks associated wit Ruthenium sublimation. La Hague Technical Department has thus developed an experimental protocol to quantify and contain the Ruthenium risk, the result of which will then be applied to Marcoule where the Ruthenium issue has appeared in recent operations. Commissioning and operating fission products evaporators is a rather standard activity on UP2 800 and UP3, while the associated experience has been decreasing in Marcoule following final shutdown in 1998. When the French atomic Energy commission decided to build and operate a new evaporator to concentrate rinsing effluents prior to vitrification in 2009, AREVA La Hague operators were mobilized to test and commission the new equipment, and train local operators. Concrete scabbling is the final stage prior to the free release of a nuclear facility. In the context of Veurey and Annecy final cleanup and declassification, large scale concrete scabbling operations were conducted, and lead to the industrialization of the process and qualification of a new process, NiThrow{sup TM} scabbling technology, developed by AREVA. This experience has now been injected into La Hague D and D scenario and has allowed a significant gain in time and cost for scabbling operations. In short, the variety of experiences and sites under the responsibility of AREVA D and D teams present significant challenges, and yet provide a unique opportunity to innovate and qualify new tools and methods which can then be shared throughout the sites. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

The Regulatory Challenges of Decommissioning Nuclear Power Plants in Korea - 13101

As of 2012, 23 units of nuclear power plants are in operation, but there is no experience of permanent shutdown and decommissioning of nuclear power plant in Korea. It is realized that, since late 1990's, improvement of the regulatory framework for decommissioning has been emphasized constantly from the point of view of International Atomic Energy Agency (IAEA)'s safety standards. And it is known that now IAEA prepare the safety requirement on decommissioning of facilities, its title is the Safe Decommissioning of Facilities, General Safety Requirement Part 6. According to the result of IAEA's Integrated Regulatory Review Service (IRRS) mission to Korea in 2011, it was recommended that the regulatory framework for decommissioning should require decommissioning plans for nuclear installations to be constructed and operated and these plans should be updated periodically. In addition, after the Fukushima nuclear disaster in Japan in March of 2011, preparedness for early decommissioning caused by an unexpected severe accident became also important issues and concerns. In this respect, it is acknowledged that the regulatory framework for decommissioning of nuclear facilities in Korea need to be improved. First of all, we identify the current status and relevant issues of regulatory framework for decommissioning of nuclear power plants compared to the IAEA's safety standards in order to achieve our goal. And then the plan is to be established for improvement of regulatory framework for decommissioning of nuclear power plants in Korea. After dealing with it, it is expected that the revised regulatory framework for decommissioning could enhance the safety regime on the decommissioning of nuclear power plants in Korea in light of international standards. (authors)

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DOE OSTI2013

Integrated Systems-Based Approach for Reaching Acceptable End Points for Groundwater - 13629

The sheer mass and nature of contaminated materials at DOE and DoD sites, makes it impractical to completely restore these sites to pre-disposal conditions. DOE faces long-term challenges, particularly with developing monitoring and end state approaches for clean-up that are protective of the environment, technically based and documented, sustainable, and most importantly cost effective. Integrated systems-based monitoring approaches (e.g., tools for characterization and monitoring, multi-component strategies, geophysical modeling) could provide novel approaches and a framework to (a) define risk-informed endpoints and/or conditions that constitute completion of cleanup and (b) provide the understanding for implementation of advanced scientific approaches to meet cleanup goals. Multi-component strategies which combine site conceptual models, biological, chemical, and physical remediation strategies, as well as iterative review and optimization have proven successful at several DOE sites. Novel tools such as enzyme probes and quantitative PCR for DNA and RNA, and innovative modeling approaches for complex subsurface environments, have been successful at facilitating the reduced operation or shutdown of pump and treat facilities and transition of clean-up activities into monitored natural attenuation remedies. Integrating novel tools with site conceptual models and other lines of evidence to characterize, optimize, and monitor long term remedial approaches for complex contaminant plumes are critical for transitioning active remediation into cost effective, yet technically defensible endpoint strategies. (authors)

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DOE OSTI2013

Present status and future plans for the restoration of Fukushima Daiichi NPS

There are many challenges at Fukushima-Daiichi Nuclear Power Station, even though the reactors have reached a condition equivalent to cold shutdown. Cooling of reactors is maintained using the circulating cooling system, which contains some temporary equipments and consists of long pipes reaching to about 4 km. The reliability of the cooling system is one of the keys to maintaining the stabilized condition. Therefore a variety of activities, such as the replacement of hoses with more durable pipes, is being conducted to improve the reliability of the cooling system. The first fuel bundle will be removed from the spent fuel pool (SFP) at Unit 4 in November, which is earlier than the original plan. In preparation for the fuel debris removal, investigations inside the primary containment vessel (PCVs) have been initiated. Visual images and other data, such as water level, radiation dose and temperature inside the PCVs, have been taken at Units 1 and 2. The issue of excessive contaminated water is the most urgent one. About 400 tons of ground water is intruding into the turbine and reactor buildings every day. This intruding water mixes with the core cooling water and becomes a large amount of contaminated water. It has to be stored on the site, despite the fact that some radioactive nuclides remain in it even after cesium is removed. The amount of water stored on the site has reached about 290,000 tons so far and is increasing. Counter measures for this issue, such as reduction of intruding water, cleaning up of the contaminated water and securing of water storage tanks, are under consideration and being conducted. However, certain recent events, such as the power loss of the SFP cooling systems and leakage of contaminated water from the underground tanks, have raised concerns about reliability. Vulnerabilities have been evaluated systematically and counter measures are being taken. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Overview of Remote Handling Equipment Used for the NPP A1 Decommissioning - 12141

The first Czechoslovak NPP A1 was in operation from 1972 to 1977 and it was finally shutdown due to an accident (level 4 according to the INES). The presence of radioactive, toxic or hazardous materials limits personnel access to facilities and therefore it is necessary to use remote handling technologies for some most difficult characterization, retrieval, decontamination and dismantling tasks. The history of remote handling technologies utilization started in nineties when the spent nuclear fuel, including those fuel assemblies damaged during the accident, was prepared for the transport to Russia. Subsequent significant development of remote handling equipment continued during implementation of the NPP A1 decommissioning project - Stage I and ongoing Stage II. Company VUJE, Inc. is the general contractor for both mentioned stages of the decommissioning project. Various remote handling manipulators and robotics arms were developed and used. It includes remotely controlled vehicle manipulator MT-15 used for characterisation tasks in hostile and radioactive environment, special robust manipulator DENAR-41 used for the decontamination of underground storage tanks and multi-purposes robotics arms MT-80 and MT-80A developed for variety of decontamination and dismantling tasks. The heavy water evaporator facility dismantling is the current task performed remotely by robotics arm MT-80. The heavy water evaporator is located inside the main production building in the room No. 220 where loose surface contamination varies from 10 Bq/cm{sup 2} to 1x10{sup 3} Bq/cm{sup 2}, dose rate is up to 1.5 mGy/h and the feeding pipeline contained liquid RAW with high tritium content. Presented manipulators have been designed for broad range of decommissioning tasks. They are used for recognition, sampling, waste retrieval from large underground tanks, decontamination and dismantling of technological equipments. Each of the mentioned fields claims specific requirements on design of manipulator, their operation and control systems as well as tools of manipulators. Precise planning of decontamination and dismantling tasks is necessary for its successful performance by remotely controlled manipulator. The example of the heavy water evaporator demonstrates typical procedure for decommissioning of contaminated technological equipment by remotely controlled manipulators - planning of decommissioning tasks, preparatory tasks, modification of applied tools and design of specific supporting constructions for manipulator and finally decontamination and dismantling themselves. Due to the particularly demanding conditions in highly contaminated A1 NPP, a team of experts with special know-how in the field of decommissioning has grown up, and unique technological equipment enabling effective and safe work in environment with a high radiation level has been developed. (authors)

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DOE OSTI2012

Waste-Incidental-to-Reprocessing Evaluation for the West Valley Demonstration Project Vitrification Melter - 12167

The Department of Energy (DOE) has determined that the vitrification melter used in the West Valley Demonstration Project can be disposed of as low-level waste (LLW) after completion of a waste-incidental-to-reprocessing evaluation performed in accordance with the evaluation process of DOE Manual 435.1-1, Radioactive Waste Management Manual. The vitrification melter - which consists of a ceramic lined, electrically heated box structure - was operated for more than 5 years melting and fusing high-level waste (HLW) slurry and glass formers and pouring the molten glass into 275 stainless steel canisters. Prior to shutdown, the melter was decontaminated by processing low-activity decontamination flush solutions and by extracting molten glass from the melter cavity. Because it could not be completely emptied, residual radioactivity conservatively estimated at approximately 170 TBq (4,600 Ci) remained in the vitrification melter. To establish whether the melter was incidental to reprocessing, DOE prepared an evaluation to demonstrate that the vitrification melter: (1) had been processed to remove key radionuclides to the maximum extent technically and economically practical; (2) would be managed to meet safety requirements comparable to the performance objectives for LLW established by the Nuclear Regulatory Commission (NRC); and (3) would be managed by DOE in accordance with DOE's requirements for LLW after it had been incorporated in a solid physical form with radionuclide concentrations that do not exceed the NRC concentration limits for Class C LLW. DOE consulted with the NRC on the draft evaluation and gave other stakeholders an opportunity to submit comments before the determination was made. The NRC submitted a request for additional information in connection with staff review of the draft evaluation; DOE provided the additional information and made improvements to the evaluation, which was issued in January 2012. DOE considered the NRC Technical Evaluation Report as well as comments received from other stakeholders prior to making its determination that the vitrification melter is not HLW, does not require permanent isolation in a geologic repository, and can be disposed of as LLW. (authors)

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DOE OSTI2012

Testing Dual Rotary Filters - 12373

The Savannah River National Laboratory (SRNL) installed and tested two hydraulically connected SpinTek{sup R} Rotary Micro-filter units to determine the behavior of a multiple filter system and develop a multi-filter automated control scheme. Developing and testing the control of multiple filters was the next step in the development of the rotary filter for deployment. The test stand was assembled using as much of the hardware planned for use in the field including instrumentation and valving. The control scheme developed will serve as the basis for the scheme used in deployment. The multi filter setup was controlled via an Emerson DeltaV control system running version 10.3 software. Emerson model MD controllers were installed to run the control algorithms developed during this test. Savannah River Remediation (SRR) Process Control Engineering personnel developed the software used to operate the process test model. While a variety of control schemes were tested, two primary algorithms provided extremely stable control as well as significant resistance to process upsets that could lead to equipment interlock conditions. The control system was tuned to provide satisfactory response to changing conditions during the operation of the multi-filter system. Stability was maintained through the startup and shutdown of one of the filter units while the second was still in operation. The equipment selected for deployment, including the concentrate discharge control valve, the pressure transmitters, and flow meters, performed well. Automation of the valve control integrated well with the control scheme and when used in concert with the other control variables, allowed automated control of the dual rotary filter system. Experience acquired on a multi-filter system behavior and with the system layout during this test helped to identify areas where the current deployment rotary filter installation design could be improved. Completion of this testing provides the necessary information on the control and system behavior that will be used in deployment on actual waste. (authors)

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DOE OSTI2012

Critical Drivers for Safety Culture: Examining Department of Energy and U.S. Army Operational Experiences - 12382

Evaluating operational incidents can provide a window into the drivers most critical to establishing and maintaining a strong safety culture, thereby minimizing the potential project risk associated with safety incidents. By examining U.S. Department of Energy (DOE) versus U.S. Army drivers in terms of regulatory and contract requirements, programs implemented to address the requirements, and example case studies of operational events, a view of the elements most critical to making a positive influence on safety culture is presented. Four case studies are used in this evaluation; two from DOE and two from U.S. Army experiences. Although the standards guiding operations at these facilities are different, there are many similarities in the level of hazards, as well as the causes and the potential consequences of the events presented. Two of the incidents examined, one from a DOE operation and the other from a U.S. Army facility, resulted in workers receiving chemical burns. The remaining two incidents are similar in that significant conduct of operations failures occurred resulting in high-level radioactive waste (in the case of the DOE facility) or chemical agent (in the case of the Army facility) being transferred outside of engineering controls. A review of the investigation reports for all four events indicates the primary causes to be failures in work planning leading to ineffective hazard evaluation and control, lack of procedure adherence, and most importantly, lack of management oversight to effectively reinforce expectations for safe work planning and execution. DOE and Army safety programs are similar, and although there are some differences in contractual requirements, the expectations for safe performance are essentially the same. This analysis concludes that instilling a positive safety culture comes down to management leadership and engagement to (1) cultivate an environment that values a questioning attitude and (2) continually reinforce expectations for the appropriate level of rigor in work planning and procedure adherence. A review of the root causes and key contributing causes to the events indicate: - Three of the four root cause analyses cite lack of management engagement (oversight, involvement, ability to recognize issues, etc.) as a root cause to the events. - Two of the four root cause analyses cite work planning failures as a root cause to the events and all cause analyses reflect work planning failures as contributing factors to the events. - All events with the exception of the Tuba City plant shutdown indicate procedure noncompliance as a key contributor; in the case of Tuba City the procedure issues were primarily related to a lack of procedures, or a lack of sufficiently detailed procedures. - All events included discussion or suggestion of a lack of a questioning attitude, either on the part of management/supervision, work planners, or workers. This analysis suggests that the most critical drivers to safety culture are: - Management engagement, - Effective work planning and procedures, and - Procedure adherence with a questioning attitude to ensure procedural problems are identified and fixed. In high-hazard operational environments the importance of robust work planning processes and procedure adherence cannot be overstated. However, having the processes by themselves is not enough. Management must actively engage in expectation setting and ensure work planning that meets expectations for hazard analysis and control, develop a culture that encourages incident reporting and a questioning attitude, and routinely observe work performance to reinforce expectations for adherence to procedures/work control documents. In conclusion, the most critical driver to achieving a workforce culture that supports safe and effective project performance can be summarized as follows: 'Management engagement to continually reinforce expectations for work planning processes and procedure adherence in an environment that cultivates a questioning attitude'. (authors)

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DOE OSTI2012

Fukushima Nuclear Crisis Recovery: A Modular Water Treatment System Deployed in Seven Weeks - 12489

On March 11, 2011, the magnitude 9.0 Great East Japan earthquake, Tohoku, hit off the Fukushima coast of Japan. This was one of the most powerful earthquakes in recorded history and the most powerful one known to have hit Japan. The ensuing tsunami devastated a huge area resulting in some 25,000 persons confirmed dead or missing. The perfect storm was complete when the tsunami then found the four reactor, Fukushima-Daiichi Nuclear Station directly in its destructive path. While recovery systems admirably survived the powerful earthquake, the seawater from the tsunami knocked the emergency cooling systems out and did extensive damage to the plant and site. Subsequent hydrogen generation caused explosions which extended this damage to a new level and further flooded the buildings with highly contaminated water. Some 2 million people were evacuated from a fifty mile radius of the area and evaluation and cleanup began. Teams were assembled in Tokyo the first week of April to lay out potential plans for the immediate treatment of some 63 million gallons (a number which later exceeded 110 million gallons) of highly contaminated water to avoid overflow from the buildings as well as supply the desperately needed clean cooling water for the reactors. A system had to be deployed with a very brief cold shake down and hot startup before the rainy season started in early June. Joined by team members Toshiba (oil removal system), AREVA (chemical precipitation system) and Hitachi-GE (RO system), Kurion (cesium removal system following the oil separator) proposed, designed, fabricated, delivered and started up a one of a kind treatment skid and over 100 metric tons of specially engineered and modified Ion Specific Media (ISM) customized for this very challenging seawater/oil application, all in seven weeks. After a very short cold shake down, the system went into operation on June 17, 2011 on actual waste waters far exceeding 1 million Bq/mL in cesium and many other isotopes. One must remember that, in addition to attempting to do isotope removal in the competition of seawater (as high as 18,000 ppm sodium due to concentration), some 350,000 gallons of turbine oil was dispersed into the flooded buildings as well. The proposed system consisted of a 4 guard vessel skid for the oil and debris, 4 skids containing 16 cesium towers in a lead-lag layout with removable vessels (sent to an interim storage facility), and a 4 polishing vessel skid for iodine removal and trace cesium levels. At a flow rate of at least 220 gallons per minute, the system has routinely removed over 99% of the cesium, the main component of the activity, since going on line. To date, some 50% of the original activity has been removed and stabilized and cold shutdown of the plant was announced on December 10, 2011. In March and April alone, 10 cubic feet of Engineered Herschelite was shipped to Seabrook Nuclear Power Plant, NPP, to support the April 1, 2011 outage cleanup; 400 cubic feet was shipped to Oak Ridge National Laboratory (ORNL) for strontium (Sr-90) ground water remediation; and 6000 cubic feet (100 metric tons, MT, or 220,400 pounds) was readied for the Fukushima Nuclear Power Station with an additional 100 MT on standby for replacement vessels. This experience and accelerated media production in the U.S. bore direct application to what was to soon be used in Fukushima. How such a sophisticated and totally unique system and huge amount of media could be deployable in such a challenging and changing matrix, and in only seven weeks, is outlined in this paper as well as the system and operation itself. As demonstrated herein, all ten major steps leading up to the readiness and acceptance of a modular emergency technology recovery system were met and in a very short period of time, thus utilizing three decades of experience to produce and deliver such a system literally in seven weeks: - EPRI - U.S. Testing and Experience Leading to Introduction to EPRI - Japan and Subsequently TEPCO Emergency Meetings - Three Mile Island (TMI) Media and Vitrification Experience by PNNL - Commercial Nuclear Power Plant Media Experience (including long term Cs removal) - DOE Low Active Waste (LAW) and High Level Waste (HLW) in High Salt and pH Conditions Media and Vitrification Experience - National Laboratory (e.g. Oak Ridge National Laboratory, ORNL) Ground Water Media Experience - Gulf Oil Spill Media Experience in Seawater - All Media Had to be Fully Tested at High Rad Levels in Seawater and Oil Before Arriving in Japan - Final Waste Form and Disposal Experience (e.g., vitrification) - 100 Metric Tons (6000 cubic feet or 220,400 pounds) of Media had to be Immediately Available with the same amount in production as replacement media. [To date, for 2011, 400 MT of media have been prepared for Japan alone.] - Remote Operation, Modular Water Treatment Equipment Design and Fabrication in both Commercial NPP and DOE Canyon Operations. (authors)

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DOE OSTI2012

US Spent (Used) Fuel Status, Management and Likely Directions- 12522

As of 2010, the US has accumulated 65,200 MTU (42,300 MTU of PWR's; 23,000 MTU of BWR's) of spent (irradiated or used) fuel from 104 operating commercial nuclear power plants situated at 65 sites in 31 States and from previously shutdown commercial nuclear power plants. Further, the Department of Energy (DOE) has responsibility for an additional 2458 MTU of DOE-owned defense and non defense spent fuel from naval nuclear power reactors, various non-commercial test reactors and reactor demonstrations. The US has no centralized large spent fuel storage facility for either commercial spent fuel or DOE-owned spent fuel. The 65,200 MTU of US spent fuel is being safely stored by US utilities at numerous reactor sites in (wet) pools or (dry) metal or concrete casks. As of November 2010, the US had 63 'independent spent fuel storage installations' (or ISFSI's) licensed by the US Nuclear Regulatory Commission located at 57 sites in 33 states. Over 1400 casks loaded with spent fuel for dry storage are at these licensed ISFSI's; 47 sites are located at commercial reactor sites and 10 are located 'away' from a reactor (AFR's) site. DOE's small fraction of a 2458 MTU spent fuel inventory, which is not commercial spent fuel, is with the exception of 2 MTU, being stored at 4 sites in 4 States. The decades old US policy of a 'once through' fuel cycle with no recycle of spent fuel was set into a state of 'mass confusion or disruption' when the new US President Obama's administration started in early 2010 stopping the only US geologic disposal repository at the Yucca Mountain site in the State of Nevada from being developed and licensed. The practical result is that US nuclear power plant operators will have to continue to be responsible for managing and storing their own spent fuel for an indefinite period of time at many different sites in order to continue to generate electricity because there is no current US government plan, schedule or policy for taking possession of accumulated spent fuel from the utilities. There are technical solutions for continuing the safe storage of spent fuel for 100 years or more and these solutions will be implemented by the US utilities that need to keep their nuclear power plants operating while the unknown political events are played out to establish future US policy decisions that can remain in place long enough regarding accumulated spent fuel inventories to implement any new US spent fuel centralized storage or disposition policy by the US government. (author)

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DOE OSTI2012

Environmental Remediation Activities in Japan Following the Fukushima Dai-ichi Reactor Incident - 12603

In March 2011, the Fukushima Dai-ichi reactor power plant was crippled by the Great Pacific earthquake and subsequent tsunami. Much of the focus in the news was on the reactor site itself as the utility company (TEPCO), the Japanese government, and experts from around the world worked to bring the damaged plants into a safe shutdown condition and stem the release of radioactivity to the environment. Most of the radioactivity released was carried out to sea with the prevailing winds. Still, as weather patterns changed and winds shifted, a significant plume of radioactive materials released from the plant deposited in the environment surrounding the plant, contaminating large land areas of the Fukushima Prefecture. The magnitude of the radiological impact to the surrounding environmental is so large that the Japanese government has had to reevaluate the meaning of 'acceptably clean'. In many respects, 'acceptably clean' cannot be a one-size-fits-all standard. The economics costs of such an approach would make impossible what is already an enormous and costly environmental response and remediation task. Thus, the Japanese government has embarked upon an approach that is both situation-specific and reasonably achievable. For example, the determination of acceptably clean for a nursery school or kindergarten play yard may be different from that for a parking lot. The acceptably clean level of residual radioactivity in the surface soil of a rice paddy is different from that in a forested area. The recognized exposure situation (scenario) thus plays a large role in the decision process. While sometimes complicated to grasp or implement, such an approach does prioritize national resources to address environment remediation based upon immediate and significant risks. In addition, the Japanese government is testing means and methods, including advanced or promising technologies, that could be proven to be effective in reducing the amount of radioactivity in the environment beyond a fixed, concentration based limit. Essentially, the definition of acceptably clean includes the concept of reasonably achievable, given the available technology, means and methods, and the cost to implement such. The Japanese government recently issued three technology demonstration contracts expressly designed to test and evaluate the available technologies, means, and methods, which, if implemented, might produce the greatest risk reduction from environmental contamination for the best value. One of the Japanese contract holders, Obayashi JV, has teamed with AMEC to demonstrate the applicability and capabilities of the Orion ScanPlot{sup SM} and ScanSort{sup SM} technologies in radiologically impacted towns both inside and immediately outside the 20 km restricted zone. This presentation provides some unique images and informative insight into the environmental radioactive impacts in and around the exclusion zone. It will provide a look at one element of the Japanese government?s efforts to achieve the greatest risk reduction that is reasonably achievable. The Orion ScanPlot{sup SM} and ScanSort{sup SM} are being used with success on the Japan Town Demonstration Project to assess pre-remedial action contamination levels, document the post-remedial action contamination levels and to precisely measure and segregate excavated soils based on their radioactive content and the prescribed segregation limits (DCS). Initial results suggest that these technologies could provide capabilities to the remedial action efforts that would result in considerable improvements in field data certainty and compliance with remedial objectives while reducing overall costs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Design and Construction of Large Size Micromegas Chambers for the ATLAS Upgrade of the Muon Spectrometer

Large area Micromegas detectors will be employed for the first time in high-energy physics experiments. A total surface of about 150 m{sup 2} of the forward regions of the Muon Spectrometer of the ATLAS detector at LHC will be equipped with 8-layer Micromegas modules. Each module extends over a surface from 2 to 3 m{sup 2} for a total active area of 1200 m{sup 2}. Together with the small strip Thin Gap Chambers they will compose the two New Small Wheels, which will replace the innermost stations of the ATLAS end-cap muon tracking system in the 2018/19 shutdown. In order to achieve a 15% transverse momentum resolution for 1 TeV muons, in addition to an excellent intrinsic resolution, the mechanical precision of each plane of the assembled module must be as good as 30 μm along the precision coordinate and 80 μm perpendicular to the chamber. In the prototyping towards the final configuration two similar quadruplets with dimensions 1.2 x 0.5 m{sup 2} have been built with the same structure as foreseen for the NSW upgrade. It represents the first example of a Micromegas quadruplet ever built, realized using the resistive-strip technology and decoupling the amplification mesh from the readout structure. All readout planes are segmented into strips with a pitch of 400 μm for a total of 4096 strips. In two of the four planes the strips are inclined by 1.5 deg. and provide a measurement of the second coordinate. The design and construction procedure of the Micromegas modules will be presented, as well as the design for the assembly of modules onto the New Small Wheel. Emphasis will be given on the methods developed to achieve the challenging mechanical precision. Measurements of deformation on chamber prototypes as a function of thermal gradients, gas over-pressure and internal stress (mesh tension and module fixation on supports) will be also shown in comparison to simulation. These tests were essential in the development of the final design in order to minimize the effects of deformations. During installation and operation all deformations and relative misalignments will be monitored through an optical alignment system and compensated in the tracking software. The optical alignment concept will also be outlined. 2 (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

The Acoustic Emission signal acquired by the microphones placed in the CABRI test device along the fourteen last R.I.A. experiments: an example of reproducible research in nuclear science

The CABRI facility is an experimental nuclear reactor of the French Atomic Energy Commission. It is located at the Cadarache Research Centre in southern France and it is designed to act as a support to the French nuclear infrastructure. The purpose of the new testing programme termed, 'CABRI International Programme' (CIP) is to study the behaviour of PWR-type fuel rods at high burnup, equipped with an 'advanced' cladding, under Reactivity Initiated Accident (RIA) conditions (such as the scenario of a control rod ejection). Within the framework of this programme, piloted and funded by the French Institute of Nuclear Radioprotection and Safety (IRSN), ten tests are to be conducted with a frequency of two tests per year. The LPRE laboratory of the CEA which is in charge of the Preparation, realisation and breakdown of the test results studies the possibility to set up a new test analysis based on the processing of signals coming from sensors placed within the test equipment. During the experimental phase, the behaviour of the fuel element generates acoustic waves which can be detected by two microphones placed upstream and downstream of the test device. Studies showed the interest to realize temporal and spectral analyses on these signals by showing the existence of signatures which can be correlated with physical phenomena as the rod failure or the test shutdown (i.e. scram). The work presented in this article results from the will to consolidate these studies. Since the main phenomenon to be tracked is the fuel rod failure, the aim would be to highlight specific events which would have been precursors of the rod failure in order to use in the future these signals for further interpretation. Such an antecedent works resumption leads to a better understanding of the experimental needs and constitutes a good initial state to prepare the new very fast digital data acquisition systems. Although all the raw data are accessible in the form of text files, analyses and graphics representations were not straightforward to reproduce from the ancient studies since that, on one hand, people who were in charge of the original work left the laboratory and on the other hand because it is not easy when the time passes, even with our own work, to be able to remember the steps of data manipulations and the exact setup: - During the ancient experiments the use of analog data acquisition systems required to digitize tapes to be able to realize computer treatments. That had had for consequence to lose the initial dating. This one must be correctly edited to do temporal comparisons. - Analyses require functions for calculations whose parameters has to be well-known to reach the same results. We thus wished to manage our workflow in the idea that it can be easily reproducible on all the experiments. The object of the work presented in this article was to put in practice this strong bind between the data, treatments and generation of the document in order not to hesitate to do the iteration principle in action. We do not have to be afraid by the data driven analyses. According to the philosophy of the literate programming, the text of the technical document is woven with the computer code that produces all the printed output as tables, graphs for the study eliminating hence the unrealistic cut and paste. This difficulty is not specific to the nuclear domain. For many years, researchers have been worked out solutions to this mundane issue. And, presently, new technologies and high-level programming languages offer us actual answers. We will firstly present the tools applied in our laboratory to implement this workflow, then we will describe the global perception carried out to continue the study of the Acoustic Emission signals recorded by the two microphones during the fourteen last CABRI R.I.A. test.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗