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At least 1,765 records · Page 98

Selection of materials for sodium fast reactor steam generators

Sodium Fast Reactor (SFR) is considered in France as the most mature technology of the different Generation IV systems. In the short-term the designing work is focused on the identification of the potential tracks to demonstrate licensing capability, availability, in-service inspection capability and economical performance. In that frame materials selection for the major components, as the steam generator, is a particularly key point managed within a French Research and Development program launched by AREVA, CEA and EDF. The choice of the material for the steam generator is indeed complex because various aspects shall be considered like mechanical and thermal properties at high temperature, interaction with sodium on one side and water and steam on the other side, resistance to wastage, procurement, fabrication, weldability and ability for inspection and in-situ intervention. The following relevant options are evaluated: the modified 9Cr1Mo ferritic-martensitic grade and the Alloy 800 austenitic grade. The objective of this paper is to assess for both candidates their abilities to reach the current SFR needs regarding material design data, from AFCEN RCC-MRx Code in particular, compatibility with environments and manufacturability. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Detection Of Special Nuclear Materials Tagged Neutrons

In order to detect Special Nuclear Materials (SNM) in unattended luggage or cargo containers in the field of homeland security, fissions are induced by 14 MeV neutrons produced by an associated particle DT neutron generator, and prompt fission particles correlated with tagged neutron are detected by plastic scintillators. SMN produce high multiplicity events due to induced fissions, whereas nonnuclear materials produce low multiplicity events due to cross-talk, (n,2n) or (n,n'γ) reactions. The data acquisition electronics is made of compact FPGA boards. The coincidence window is triggered by the alpha particle detection, allowing to tag the emission date and direction of the 14 MeV interrogating neutron. The first part of the paper presents experiment vs. calculation comparisons to validate MCNP-PoliMi simulations and the post-processing tools developed with the data analysis framework ROOT. Measurements have been performed using different targets (iron, lead, graphite), first with small plastic scintillators (10 x 10 x 10 cm{sup 3}) and then with large detectors (10 x 10 x 100 cm{sup 3}) to demonstrate that nuclear materials can be differentiated from nonnuclear dense materials (iron, lead) in iron and wood matrixes. A special attention is paid on SNM detection in abandoned luggage. In the second part of the paper, the performances of a cargo container inspection system are studied by numerical simulation, following previous work reported in. Detectors dimensions and shielding against the neutron generator background are optimized for container inspection. Events not correlated to an alpha particle (uncorrelated background), counting statistics, time and energy resolutions of the data acquisition system are all taken into account in a realistic numerical model. The impact of the container matrix (iron, ceramic, wood) has been investigated by studying the system capability to detect a few kilograms of SNM in different positions in the cargo container, within 10 min acquisitions. (authors)

36 MATERIALS SCIENCE↗

Robotics Technologies for Decommissioning of Fukushima Daiichi - 17077

The Fukushima Daiichi Nuclear Power Station (NPS) of Tokyo Electric Power Company (TEPCO) was the site of a severe nuclear accident after the 2011 Off the Pacific coast of Tohoku Earthquake and subsequent tsunami. Extensive structural damage occurred in the NPS facilities, including such environments as the reactor buildings and the primary containment vessels (PCVs). These are high radiation areas, having underwater areas, narrow spaces and so on. It is necessary to determine the extent of contamination and the location of debris to plan decommissioning operations. Survey targets in the past few years have been focused on inspections of the basement of the reactor buildings and the inside of the PCVs. Human operators cannot access these environments and work in them directly. Therefore, remotely operated robots and measurement technologies are needed to carry out various operations. For underwater surveys, we have developed a submersible crawling swimming robot that can avoid obstacles and inspect wide areas as it moves around. A shape-changing robot has been developed for passing through narrow spaces and crawling on rough floors. An ultrasonic imaging method combined with convex scanning and a divergent beam has been proposed to visualize environments in turbid water. In this paper, we describe details of the two survey robots and the underwater imaging method. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Current Problems and Solutions for the Long-term Storage of LLW in Germany - 17512

The primary aim of the Lower Saxony regulating authority is that all radioactive waste in long-term interim storage is made safe and their contents precisely documented. The requirements for regulating authority include: documentation, recovery, inspection, reconditioning, longer-term interim storage with full accessibility for inspection and transfer to the national disposal facility. The waste acceptance requirements of the disposal site Konrad as well as requirements for the interim storage facilities have to be met for all stored waste packages. Beside this, some challenges came up in the last few years, e.g. waste products with restrictions concerning the stability because of corrosion, fouling or fermentation processes. Some of these examples are shown. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Post-closure program streamlining at the Nevada national security site - 15290

More than 200 sites on the Nevada National Security Site (NNSS), Nevada Test and Training Range (NTTR), and Tonopah Test Range (TTR) have been closed in place. At these sites, contamination has been left in place, and protection of human health, safety, and the environment is ensured through the establishment of use restrictions (URs). Post-closure requirements were initially established in the Closure Report (CR) for each site and include a variety of inspection, monitoring, maintenance, and reporting commitments. Post-closure requirements and site controls (fences, monuments, soil covers, and postings) may vary depending on site location, accessibility, land use scenario, and the nature of contaminants left in place. Corrective Action Sites (CASs) on the NNSS, NTTR, and TTR have been identified for closure under the Federal Facility Agreement and Consent Order (FFACO), which was signed in 1996. Over 1,900 CASs have been closed under the framework of the FFACO, and more than 200 of those have URs. Many post-closure requirements have been modified over time. In addition, because the scope of the program has spanned almost 20 years, the post-closure requirements have varied due to changing strategies and expectations. Because of the number of sites closed over the course of the program, post-closure requirements are found in hundreds of closure documents, addendums, errata sheets, Records of Technical Change, inspection reports, permit modifications, and correspondence. Furthermore, resources such as personnel and funding necessary to conduct the post-closure program have diminished, while computer capabilities and configuration control principles have improved since closure activities under the FFACO began. Because of these factors, to effectively manage the post-closure program, ensure that all commitments are being met, and improve the ease of identifying the most current post-closure requirements for each UR, the current requirements for every use-restricted site under the purview of the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office will be consolidated into a single Post-Closure Plan. The Consolidated Post-Closure Plan will also outline a strategy and present possible criteria that may be used for future evaluation of post-closure requirements. Continual assessment of the post-closure program will ensure that the post-closure requirements for URs across the NNSS, NTTR, and TTR are consistently justifiable and sufficient to protect site workers, the public, and the environment while increasing efficiency. Future changes to post-closure requirements and justification for those changes will be recorded in revisions to the Consolidated Post-Closure Plan. This process will provide a comprehensive, consistent, and streamlined method for evaluation of post-closure requirements and implementation of changes. (authors)

54 ENVIRONMENTAL SCIENCES↗

Operational plant lifetime assessment - UK national nuclear laboratory experience - 15366

Inspection and monitoring of the condition of key plant assets at the Sellafield Site is an important task in helping to enable the continued operation of the reprocessing and waste management facilities. This takes place within a range of environments, from non-active areas to highly active black cells where no man-access is possible. NNL staff work closely with the plant operators and other contractors in carrying out inspection and monitoring tasks, and have a particular expertise in the black cell environments where increased challenges exist. As plants age more work is required in order to underpin ongoing operations and understand the operational lifetime of key assets in order to achieve maximum throughput and utilisation. This paper discusses how expertise in a variety of different technical disciplines is combined in order to both understand and extend the operational life of the Highly Active (HA) evaporators at Sellafield, and a small number of further similar examples are also discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Hanford double-shell tank extent-of-condition construction review - 15498

During routine visual inspections of Hanford double-shell waste tank 241-AY-102 (AY-102), anomalies were identified on the annulus floor which resulted in further evaluations. Following a formal leak assessment in October 2012, Washington River Protection Solutions, LLC (WRPS) determined that the primary tank of AY-102 was leaking. A formal leak assessment, documented in RPP-ASMT-53793, Tank 241-AY-102 Leak Assessment Report, identified first-of-a-kind construction difficulties and trial-and-error repairs as major contributing factors to tank failure. To determine if improvements in double-shell tank (DST) construction occurred after construction of tank AY-102, a detailed review and evaluation of historical construction records was performed for Hanford's remaining twenty-seven DSTs. Review involved research of 241 boxes of historical project documentation to better understand the condition of the Hanford DST farms, noting similarities in construction difficulties/issues to tank AY-102. Information gathered provides valuable insight regarding construction difficulties, future tank operations decisions, and guidance of the current tank inspection program. Should new waste storage tanks be constructed in the future, these reviews also provide valuable lessons-learned.(authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NUHOMS{sup R} EOS (Extended Optimized Storage): Robust, Advanced, and Dual Purpose Dry Storage System - 16472

AREVA TN has completed the design, analysis, and US NRC license submittal for the NUHOMS{sup R} Extended Optimized Storage (EOS) system. NUHOMS{sup R} EOS offers a large capacity and high heat load dry storage and transportation solution for UNF. EOS is not a new system, but an evolution of the already-proven safe and low-risk horizontal NUHOMS{sup R} storage system, which is one of the most widely used UNF dry storage systems in the United States. The EOS system features a 37PTH dry shielded canister (DSC) for 37 PWR UNF assemblies and an 89BTH DSC for the 89 BWR UNF assemblies. The EOS DSCs will be stored inside a horizontal storage module (HSM) that is a robust, fortress-like concrete storage overpack. Designed with the lessons learned from the Fukushima event in Japan in mind, EOS has robust margins for beyond design basis (BDB) events and offers the lowest risk and highest safety margin of any dry storage system design for recovery from BDB events. EOS also offers innovative aging management features with an aging management program (AMP) to assure safe and low dose long-term storage at the utility sites. EOS incorporates ports for AMP inspections and offers the option of using duplex stainless steel (DSS) for the DSC shell in order to negate corrosion issues. EOS is stored horizontally and above ground, making it easy and quick to access for visual inspection, monitoring, and mitigating actions during AMP. NUHOMS{sup R} EOS is optimized for high capacity, high burnup, high enrichments, and very low cooling times for UNF assemblies. It enables the highest heat load per canister (50 kW) and low dose rates. Even with high heat loads, EOS has significant margins to design basis limits. As the future of used fuel disposition will require the need to move fuel off-site to interim or consolidated storage, EOS is the ideal solution as it enables on-site loading and unloading operations that mitigate risk when transferring fuel on-site or transporting fuel off-site. The EOS design can be safely located in a traditional outdoor setting with fortress-like concrete reinforced modules that offer superior resistance to attack or can be safely operated inside of a building enclosure for those utilities desiring a more discreet siting for its used fuel storage systems. EOS has an innovative basket design that is made of high strength low alloy (HSLA) material and is easy to fabricate due to its non-welded design. Its innovative inlet and outlet air vent design, which incorporates dose reduction hardware, allows for optimal ambient airflow for highly efficient heat transfer with minimal radiation dose. The NUHOMS{sup R} EOS storage modules are interlocked in a side-by-side configuration offering superior and robust seismic performance. The license application for NUHOMS{sup R} EOS is currently under US NRC review and this system will be available for deployment in 2017. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Regulation of the Shutdown Defueling and Decommissioning Reactors in the United Kingdom - 16576

Nuclear policy, including for decommissioning, in the United Kingdom is set by the UK Government. The Nuclear Decommissioning Authority, NDA, is responsible for managing the clean-up of the UK's nuclear legacy which includes the first generation of Magnox power stations. Magnox Ltd (MXL), owned by Cavendish Fluor Partnership, is the licensee and the management and operations contractor responsible for ten Magnox reactor sites and two restoration sites in the UK, working for the sites' owner, NDA. The work of the company is diverse covering electricity generation at Wylfa (until December 2015), defueling, and preparation for an interim end state, known as Care and Maintenance (C and M) which will precede the final site clearance. The C and M phase is expected to last over 75 years during which the sites are maintained in a passively safe and secure state while radiation levels are left to decay naturally. Preparations for this phase include removing fuel from the reactors and storage ponds; retrieving intermediate level waste from current stores and treating it or converting it to forms suitable for long-term storage or disposal and making sure of the integrity of critical safety systems of the de-fuelled reactors. The Decommissioning, Fuel and Waste programme within the Office for Nuclear Regulation, ONR, regulates nuclear safety in the 12 Magnox nuclear licensed sites. ONR works with the environmental regulators who are responsible for environmental protection on nuclear sites. The focus of this paper is the regulation of nuclear safety by ONR across the Magnox sites and the interactions with nuclear security and the environmental aspects required to achieve effective, proportionate and efficient regulation of these sites in the UK. ONR's strategy for regulating the Magnox defueling and decommissioning sites is focused on enabling the licensee to complete the initial stages of decommissioning and reduce the nuclear and radioactive hazards on its sites. To this effect, ONR's revised inspection guidance is centred on inspecting the (remaining) important systems for safety identified in the safety cases. Other interventions focus on engaging with the licensee on safety significant decommissioning projects to gain evidence of the licensee's readiness prior to and during these activities rather than formal assessment of the associated safety cases. In light of reduced hazards and risks across the de-fueled and decommissioning Magnox reactors, ONR's revised regulatory expectation regarding evolution of emergency arrangements during decommissioning is based on key transition points that follow electricity generation. Implementation of these revised strategies as well as organisational restructuring within the team responsible for regulation of the Magnox fleet has resulted in improved efficiency and consistency in regulatory approach whilst securing the required safety standards in managing decommissioning activities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Autonomous 3D and Radiation Mapping in Tunnel Environments Using Aerial Robots - 18156

In this paper we present the algorithms and systems developed to enable autonomous 3D and radiation mapping, as well as radiological source localization in tunnel environments using aerial robots. The developed technologies aim to address the need of autonomous robotic inspection of nuclear sites such as facilities in the process of being decommissioned. As a motivational example, the PUREX tunnels at the Hanford Site are considered and the robot is designed to be able to conduct multi-modal 3D/radiation characterization of the tunnel environment without any prior knowledge or map provided beforehand. In order to enable the necessary GPS-denied simultaneous localization and mapping in dark, visually-degraded tunnel environments, a multi-modal mapping unit was developed which fuses the data from stereo vision cameras synchronized with flashing LEDs, inertial sensors, and three time-of-flight 3D depth sensors. The robot's pose and surrounding map are estimated at 20 Hz through a depth-enhanced visual- inertial localization and mapping framework. Provided this ability to estimate the robot's pose and develop a dense map in real-time, a receding horizon sampling-based autonomous exploration path planning algorithm was utilized and equips the robot with the ability to explore its environment given no prior map. The aforementioned key robotic functionalities are further extended through the integration of miniaturized fully solid-state scintillation detectors, utilizing advanced scintillators and Silicon Photomultipliers. The derived sensor data is employed for radiation mapping, which when fused with the 3D maps and the spectroscopic source localization methods, allow the aerial robotic platform to localize the radiation source and identify the specific source isotope. The source location is estimated through a geometric algorithm which can localize a source given only three measurements and short dwell times. The developed systems were tested using two small aerial robot platforms of the F450 and F550 frame classes with weights of 2.6 kg and 2.9 kg respectively. All localization and planning as well as high-level control algorithms are executed onboard. The scintillation detector is integrated on the robot and its data is collected through a dedicated microcontroller. A set of field experiments were conducted inside the dark and GPS-denied environment of a 4.25 m wide, 7.5 m tall, and 137 m long railroad tunnel. A Cs-137 source was placed within the tunnel, the associated 3D/radiation maps were derived, and the source location was estimated. Through these efforts, the goal is to develop the robotic technology that would enable the autonomous systematic inspection or emergency response in nuclear waste decommissioning facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Assessment of Quality Management for Transport Packages Not Requiring Authority Design Approval - 18194

The majority of transports of radioactive materials are carried out in packages which don't need a package design approval of a competent authority. Low active radioactive materials are transported in such kind of packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. In Germany the decision to phase out nuclear energy leads to a strong demand for packages to transport low and middle active radioactive waste due to the dismantling and decommissioning of nuclear power plants. According to IAEA regulations the 'non-competent authority approved package types' are the excepted packages and the industrial packages of Type IP-1, IP-2 and IP-3 and of Type A. For the packages of Type IP-2, IP-3 and Type A an assessment by the German competent authority is required for the quality management for the design, manufacture, testing, documentation, use, maintenance and inspection. In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the regulatory density for packages requiring competent authority design approval. The countries apply different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. In order to help manufactures, which measurements and documents are needed for the assessment procedure, BAM developed a guideline. The guideline BAM-GGR 016 [5] describes the needed quality management for the design, manufacture, testing, documentation, use, maintenance and inspection of a packaging and the package. The poster to be presented will explain the different requirements of the quality assurance measures and give examples. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Augmented Reality Tools for Nuclear Facilities - 18318

In this work we explore the possibility of improving nuclear facility operations through the use of emerging augmented reality technology. Specifically we focus on the use of augmented reality headsets. Augmented reality opens up the possibility for easy, on-the-fly access to documentation related to canisters, rooms, and carts. It also can be used to help plan complicated material moves in ways that do not violate nuclear criticality safety regulations. Modern augmented reality technology can also be interacted with using gaze, gesture and voice commands, thus making it particularly attractive for facilitating the documentation of nuclear material operations and work verification. An augmented reality-based nuclear facility criticality safety system would have a number of benefits for operations in nuclear facilities. This work could be used across the DOE complex to ensure technicians in gloveboxes work safely and securely. It could greatly reduce downtime associated with administrative criticality incidents. It would be a step forward in the process of improving the quality of documentation of the work that occurs in nuclear facilities while simultaneously increasing the ease with which this documentation can be collected. The augmented reality effectively endows the glovebox technicians with a sense for administrative nuclear criticality safety and the reactivity changes caused directly by their actions. Augmented reality also has great potential for facilitating inspections of nuclear infrastructure. Examples of Augmented Reality applications that facilitate infrastructure inspection and work verification will also be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Highlights from the Postirradiation Examination of AFC-3A and AFC-3B

Highlights from the postirradiation examination of the AFC-3A and AFC-3B irradiation test are presented. Baseline non-destructive and destructive examinations of these tests have been completed at this time. Results from visual examination, neutron radiography, gamma spectrometry, dimensional inspection, fission gas release, chemical burnup analysis, and optical microscopy have been collected. All the rodlets from these irradiations appear to have performed well during irradiation. Baseline PIE has been completed on rodlets from AFC- 3A, and AFC-3B. This includes visual exams, neutron radiography, gamma spectrometry. dimensional inspection, fission gas release evaluation, optical microscopy. microhardness evaluation. and chemical analysis. The results compiled here are a sampling of the full results. Baseline PIE provides an engineering or macroscopic scale evaluation of fuel performance and prepares samples for a more detailed phenomenological study of observed performance. Advanced PIE techniques are desired to obtain microscopic elemental and structural information both in the fuel, in the cladding, and in the interface where the composition of the FCCI layer is of great interest. The AFC-3 tests do appear to have performed roughly as designed during the irradiation. The FCCI behavior of Mo alloy metallic fuel versus Zr alloy metallic fuel has been confirmed. Results from this irradiation such as the FCCI observed in AFC-3A R1 and AFC-3A R2 can be used to eliminate Mo alloys as a choice in future irradiations where ultra-high burnup is a stated goal. Annular fuels are a preferred method for many high burnup core designs. The redistribution behavior of the annular U-Zr alloys should be studied further to better understand how this might impact high burnup annular fuel performance. The thermal bonding between annular fuel and cladding should also be studied further as there appear to be localized areas of FCCI forming in both AFC-3A R4 and AFC-3B R2, while no FCCI was observed in the Zr-based Na-bonded or Mo-based low smear density Na-bonded fuel. The performance of the low smear density. solid fuels is less clear. The neutron radiography and optical microscopy cross sections suggest the fuel is not forming a radially symmetric structure which may be causing undesirable localized temperature variations. Further study both by electron microscopy and by simulation to evaluate the implications of the non-symmetric fuel form are needed to better understand whether this fuel form merits further study under irradiation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Inventory and Characterization of Categories 3-5 Disused Sealed Radioactive Sources - 19284

Brazil has one of the biggest inventories of SRS in the world, not only because of the large industrial park, the well-developed health care infrastructure and the numerous mining operations, but also by the use of radioactive sources in consumer products, such as smoke detectors, lightning arresters, surge protectors and even cardiac pacemakers. Presently, the inventory of DSRS at the WML is of more than 10,000; if the sources from lightning arresters, smoke detectors and Ir-192 with activities above the clearance levels are included, the estimate is over 150,000. The Waste Management Laboratory (WML) of the Nuclear and Energy Research Institute (IPEN/CNEN-SP) is responsible for the management and treatment of these sources. An important aspect to consider is that Brazil does not have a defined strategy or a protocol for screening and inventorying these sources, whose management was limited only to storage. The WML started the development of methods and operational work, to characterize and reduce the total volume of this inventory for the borehole-type final disposal facility yet to be constructed, where the DSRS will be disposed of. The information available from these sources is in a database created in 1980, when the sources began to be received. Several records on the FRSD that were checked are incomplete or simply incorrect, since there is no correspondence between the dates of entry of these sources in the storage and their physical characteristics. Some reasons can be listed to try to understand the divergences: the inspection when receiving the sources was not possible, no inspection procedure was defined, there were technical difficulties to open the packages. The objective of this work is to organize the inventory in a standardized way, control access, traceability, reduction of the necessary time to locate and characterize the sources upon reception, and of the repetitive work in the operational routine, and finally release of space in the storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Safeguards Lessons Learned from the Rokkasho Reprocessing Plant - 19502

Once operational, the Rokkasho Reprocessing Plant (RRP) will have an operating throughput of 800 MT of uranium in irradiated LWR fuel per year. In order to establish effective and efficient safeguards at RRP, the safeguards discussion with the IAEA was initiated in 1988. However, at that time there were no available models or guidelines that could be used as a reference and multiple difficulties were recognized for the implementation of safeguards into a large scale reprocessing plant. Therefore, from 1988 to 1992, a multinational forum, referred to as LASCAR (Large Scale Reprocessing Plant Safeguards) [1], provided the recommendations on how an effective safeguards approach could be implemented into a large scale reprocessing plant while efficiently using existing resources. During the course of safeguards discussions between the IAEA, the State, and JNFL, the use of operator's equipment and measurement systems (by signal splitting with authentication) for inspections was studied. JNFL also examined incorporating inspection requirements into the process design: to reduce the costs associated with system installation and future system maintenance and to ensure effective, efficient, and credible safeguards implementation into the RRP (this approach has essentially become the 'Safeguards by Design (SBD)' process). From these studies and JNFL's experience of implementing safeguards into a new plant [2], JNFL established and applied signal splitting from operator's equipment as a means for providing inspectors required safeguards information. In addition, the LASCAR findings were utilized to establish the safeguards implemented into the RRP. Safeguards implementation into RRP commenced March 31, 2006. On June 4, 2008, the IAEA approved the 'integrated safeguards approach' for RRP. This paper provides a high level overview of the major safeguards implemented at RRP and the lessons learned from this implementation that can be applied to existing or future similar facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A Systematic Evidence Based Performance Approach to Regulation of Nuclear Sites in England and Wales - 20058

The Environment Agency for England has developed a systematic evidence-based approach to pursue our strategic environmental objectives for the regulation of nuclear sites. We use an annual evidence review process to ensure effective and efficient targeting of limited resources to achieve those objectives. Over the last 8 years, we have been developing and refining this approach to ensure risk-based and value driven regulation. The approach comprises nuclear site and nuclear sector review processes known as Site Environment Review (SER) and Nuclear Environment Review (NER). This approach complements our regulation of nuclear site permit holders under the Environmental Permitting Regulations (EPR). We deliver our regulation of nuclear sites in England and Wales alongside the Office for Nuclear Regulation. The SER process involves the lead regulator for each nuclear site assessing the permit holder's environmental performance across 14 themes, set within the context of the site's main activities and associated waste disposals. Our themes include environmental leadership, resources and climate change, radioactive waste management, facility management and decommissioning, groundwater, and environmental radiological protection. We use evidence from site inspections and working within our subject matter groups to grade current and predicted future environmental performance. We are particularly interested in sustainability and the application of Best Available Techniques (BAT) to prevent the creation, and minimise the disposal, of radioactive wastes. We use risk analysis (strengths, weaknesses, threats and opportunities) to examine performance against our strategic environmental objectives, which are set out, in our 5-year Nuclear Delivery Plan (NDP). The output supports the targeting of our resources at each nuclear site. We consult the relevant permit holders on the SER priorities and use their feedback to refine our plans. We expect all permit holders to take account of our priorities when considering their own programmes of work, objectives and plans. The NER process brings together what we learn and achieve through regulation across the sector. It provides input to planning priorities, supported by qualitative and semi-quantitative evidence. It covers the 28 nuclear sites in England and Wales and spans the same 14 environmental themes. During the process we collate, compile and summarise evidence from the SERs and other sources such as inspection reports and evidence from our other nuclear work programmes. The output is the NER annual report. This provides a snapshot of the status of the nuclear sector and gives insights to enable us to regulate more efficiently and effectively. It also takes account of cross-cutting issues and risks such as changes in international standards, domestic policy, regulatory framework, domestic standards and guidance, learning from experience such as incidents, events and good practice, and innovation, research and development. It provides graphics that illustrate the grading of environmental performance for the nuclear sector across the fourteen environmental themes. This analysis allows benchmarking of nuclear site's environmental performance and the visualisation provides a convenient comparison of performance across themes, sites, and over time. We use this intelligence to inform our investment in training and development of our staff, and our cross-cutting engagement on strategic issues with government, the Nuclear Decommissioning Authority (NDA) and other corporate organisations. Adopting this approach can provide benefits with organisational reputation, stakeholder participation and ensuring value from the public investment. This paper describes the history of the SER/NER process, a selection of outputs from the process and ideas for improvement. The paper will be of interest to other regulators and organisations across the world that are interested in supporting continuous improvement. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Enhancing the ATR Primary Coolant System: A 3D Modeling Approach

The project consisted in system inspections to the ATR Primary Coolant System involving welds, fittings, motors, pumps, flanges, and heat exchangers to enhance Inservice Inspection program as required by DOE orders. NOTE: This article is to be published as "DOE & DOE Contractors Only" in the OPEXShare application, which means it will be available for viewing to DOE & DOE Contractor registered users only. This article can also be used by CAES for their training and safety meetings.

42 - ENGINEERING↗

Comprehensive study of performance degradation of field‐mounted photovoltaic modules in India

Abstract The All India Survey of Photovoltaic Module Reliability 2014 is an enhanced version of the survey conducted in the previous year, with detailed characterization of PV modules including current‐voltage, infrared and electroluminescence imaging, visual inspection, insulation resistance test and interconnect breakage test. More than a thousand modules were inspected in the field and the main results of the survey are presented in this paper. The average P max degradation rate for the so‐called ‘good’ modules (Group X) is 1.33%/year which is higher than that commonly projected by manufacturers, and widely employed in financial calculations. Modules falling in the ‘not‐so‐good’ category (Group Y) show even higher degradation rates, and it is at least partly due to higher number of micro‐cracks in the modules, and increased degradation of the packaging materials like encapsulant, backsheet, etc. Modules in ‘Hot’ climates degrade faster than modules in the ‘Non‐Hot’ climates. Degradation in fill factor is the primary cause for performance degradation in the young modules (ages <5 years), whereas short‐circuit current degradation is the main contributor to power degradation in the older modules. Small installations (<100 kW p capacity) show higher degradation than large systems, which may be partly due to lack of proper due diligence by the owner at the time of procurement and installation.

Dubey, Rajiv↗