Search NASA⌕ Search

SEARCH · Search NASA

Results for “OPERATION AND MAINTENANCE”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

LCOE Uncertainty Analysis for Hydropower using Monte Carlo Simulations

Levelized Cost of Energy (LCOE) is an important metric to evaluate the cost and performance of electricity production generation alternatives, and combined with other measures, can be used to assess the economics of future hydropower development. Multiple assumptions on input parameters are required to calculate the LCOE, which each contain some level of uncertainty, in turn affecting the accuracy of LCOE results. This paper explores these uncertainties, their sources, and ultimately the level of variability they introduce at the screening level of project evaluation for non-powered dams (NPDs) across the U.S.Owing to site-specific differences in site design, the LCOE for hydropower varies significantly from project to project unlike technologies with more standardized configurations such as wind and gas. Therefore, to assess the impact of LCOE input uncertainty on the economics of U.S. hydropower resources, these uncertainties must be modeled across the population of potential opportunities. To demonstrate the impact of uncertainty, resource data from a recent nationwide non-powered dam (NPD) resource assessment (Hadjerioua et al., 2012) and screening-level predictive cost equations (O’Connor et al., 2015) are used to quantify and evaluate uncertainties in project capital and operations & maintenance costs, and generation potential at broad scale. LCOE dependence on financial assumptions is also evaluated on a sensitivity basis to explore ownership/investment implications on project economics for the U.S. hydropower fleet.The results indicate that the LCOE for U.S. NPDs varies substantially. The LCOE estimates for the potential NPD projects of capacity greater than 1 MW range from 40 to 182 $/MWh, with average of 106 $/MWh. 4,000 MW could be developed through projects with individual LCOE values below 100 $/MWh. The results also indicate that typically 90 % of LCOE uncertainty can be attributed to uncertainties in capital costs and energy production; however, for small projects (below 10 MW) O&M uncertainty plays an important role. The potential NPDs have costs with uncertainty bands extending well above 100 $/MWh in some cases. A truly representative evaluation should complement LCOE with an assessment of potential revenues from the provision of ancillary services. Moreover, hydroelectric production is often part of multipurpose water resource projects where other purposes (e.g., irrigation, flood control, navigation, recreation, etc.) also provide value. Neither of these considerations are accounted for LCOE.

Chalise, Dol Raj↗
DOE OSTI2017

Solar Ready Vets Curriculum Design

The 5-week SRV program includes four sets of program learning goals aligned around (1) the NABCEP Entry Level body of knowledge; (2) gaining hands-on experience with solar system site analysis, design, installation, commissioning, operation, maintenance and financial considerations; (3) Safety issues unique to solar + OSHA 30; (4) Transition planning and individual support of entry into the solar industry. These goals, and the learning objectives associate with each, are pursued in parallel during the course.

14 SOLAR ENERGY↗
DOE OSTI2017

Industrial Energy Training and Certification

Compressed air systems, ammonia refrigeration systems, chilled water systems, steam systems, process heating systems, combined heat and power systems, pump systems and fan systems are major industrial energy systems commonly found in manufacturing facilities. Efficiency of these systems contributes significantly to whole facilities' energy performance. On the national, even international level, well-structured training and highly recognized certification programs help develop a highly-skilled and qualified workforce to maintain and improve facilities' energy performance, particularly as technologies within these systems become more advanced. The purpose of this paper is to review currently available training and certification programs focusing on these systems and to identify the gap between market's needs and currently available programs. Three major conclusions are: first, most training programs focus on operations, maintenance, safety and design although some briefly touch the energy performance aspect; second, except CRES by RETA and PSA Certificate and PSAP Master Certification by HI, no other certifications had been found emphasizing on knowledge and skills for improving and maintaining these systems' energy performance; third, developing energy efficiency focused training and ANSI accredited certification programs on these energy systems will fill the gap between market's needs and currently available programs.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2017

Industrial Energy Training and Certification

We present compressed air systems, ammonia refrigeration systems, chilled water systems, steam systems, process heating systems, combined heat and power systems, pump systems and fan systems are major industrial energy systems commonly found in manufacturing facilities. Efficiency of these systems contributes significantly to whole facilities' energy performance. On the national, even international level, well-structured training and highly recognized certification programs help develop a highly-skilled and qualified workforce to maintain and improve facilities' energy performance, particularly as technologies within these systems become more advanced. The purpose of this paper is to review currently available training and certification programs focusing on these systems and to identify the gap between market's needs and currently available programs. Three major conclusions are: first, most training programs focus on operations, maintenance, safety and design although some briefly touch the energy performance aspect; second, except CRES by RETA and PSA Certificate and PSAP Master Certification by HI, no other certifications had been found emphasizing on knowledge and skills for improving and maintaining these systems' energy performance; third, developing energy efficiency focused training and ANSI accredited certification programs on these energy systems will fill the gap between market's needs and currently available programs.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2021

Hybrid Zircex Hydrochlorinator

This project was driven by the desire to create easier and faster ways to load Advanced Test Reactor (ATR) Fuel Elements into the Hybrid Zircex hydrochlorinator. The primary requirements for this project were to minimize the potential release of radioactive contamination, minimize maintenance operations to replace or remove the lid seal, and allow for remote usage. Three preliminary designs were made. These designs allow ATR fuel elements to be loaded into the hydrochlorinator more efficiently while fulfilling the requirements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Life Cycle Assessment for Closed-Loop Pumped Hydropower Energy Storage in the United States

The federal government has initiated an aggressive set of policies to achieve a net-zero carbon emission goal for the electricity sector by 2050. As a result, rapid growth in deployment of renewable energy technologies is expected. Most commercially mature technologies are temporally variable and do not provide grid inertia, while renewable technologies with high projected deployment have intermittent generation methods. Energy storage technologies are needed to both dispatch power on-demand and help provide the needed grid inertia. Pumped storage hydro (PSH) is a well-established technology that has gained renewed interest in recent years offering energy-balancing, grid stability, control of electrical network frequency, and large-scale storage capacity. For widespread adoption of PSH, more information is needed regarding its current life cycle environmental impacts. The objective of this study is to perform a full life cycle assessment (LCA) of new closed-loop PSH in the U.S. The functional unit for this study is 1 kWh of electrical power delivered to the grid and the base case project lifetime is 80 years. The life cycle inventory for this project accounts for all material and energy flows associated with the green-field construction, operation, maintenance, and decommissioning of a closed-loop PSH plant in the U.S. Collected data represents a range of potential PSH specifications and geographic locations coming from all prospective closed-loop PSH installations in the U.S. with data available. In addition, existing PSH installations are used to provide assumptions for inventory inputs. Results presented will include the global warming potential (GWP IPCC 100a) and Energy Return on Investment (EROI) from our base case (average PSH installation) as well as from scenario analyses and model sensitivity. These results will be compared to the impacts from existing PSH sites and alternate storage technologies. Methods align with the assumptions and guidelines put in place by previous PSH LCAs to ensure an accurate comparison with the results from this report.

ENERGY PLANNING, POLICY, AND ECONOMY,HYDRO ENERGY↗
DOE OSTI2022

WBS 1.2.3.405 - Life Cycle Assessment of Storage Technologies

Recent commitments by the Biden administration have established targets to achieve a net-zero energy system by 2050. Meeting these targets will spur a rapid transition to clean energy technologies and a commensurate need to develop and deploy energy storage technologies at scale. Pumped Storage Hydro (PSH) is expected to be part of this solution because its ability to provide grid flexibility and stability and enable the dispatching of disparate variable renewable energy technologies. Despite PSH being a mature technology with a history of deployment dating back several decades, there is very little information on the greenhouse gas (GHG) implications of PSH as compared to other storage technologies. The objective of this project is to perform a full lifecycle assessment (LCA) of new PSH projects in the U.S. This LCA includes all project phases (resource extraction, construction, operation, maintenance, end-of-life). The functional unit for this study is 1 kWh electricity delivered by system to grid substation connection point and the estimated lifetime for our base case is 80 years. Data used in this study are based on over 30 potential PSH projects that are in preliminary planning phases and are represent a wide range of potential closed-loop PSH systems in terms of location, technology, and capacity. The project approach, data sources, and modeling assumptions have been informed by a technical review committee of stakeholders that include experts from academia, national and international government, industry, and utilities. The GHGs and energy return on investment (EROI) from PSH will be compared to other storage technologies (e.g., stationary battery storage). Results from this project will improve the PSH community's understanding of the environmental impacts and sustainability of new PSH projects and how PSH compares to other storage technologies. The approach used in this project relies on open-source programming. The analysis framework (source code and data) and will be made publicly available at the end of the project. In addition to reporting results for the base case, we will perform rigorous sensitivity analysis to identify the major drivers, understand impacts of different configurations, and future energy markets. Results from this project will be published in a suitable journal.

ENERGY PLANNING, POLICY, AND ECONOMY,HYDRO ENERGY↗
DOE OSTI2022

Transferring Photovoltaic Lessons Learned to Concentrating Solar Power

This presentations outlines the following: 1) Points worth noting in the history of the PV industry, 2) PV benefits from standardization, 3) PV sharing reliability challenges since the 1975, 4) PV projects from an independent engineering (IE) perspective, 5) Site selection (what NOT to do), 6) PV Operations & Maintenance over time, 7) Soiling research (remove the silos), 8) Noting a few relevant PV IEC standards, and 9) PVfleets data sharing.

bio-soiling↗
DOE OSTI2012

AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is further surrounded by a substantial 'steel concrete' composite shield building. The containment vessel is not affected by external flooding, and the shield building design provides hazard protection beyond that provided by a comparable reinforced concrete structure. The intent of this paper is to demonstrate the robustness of the AP1000 design against extreme events. The paper will focus on the plants ability to withstand extreme external events such as beyond design basis flooding, seismic events, and malicious aircraft impact. The paper will highlight the robustness of the AP1000 nuclear island design including the protection provided by the unique AP1000 composite shield building. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Response time consideration for digitalized safety instrumentation applied to research reactor

A digitalized NMS (Neutron Measurement System) has a better performance than an analogue type in aspects of operating, maintenance and calibration. But some limitations in this system were found in the response time due to the inherent digital filtering time. The response time requirement as a kind of safety requirement is a significant factor that should be met. To satisfy the requirements, two approaches are herein suggested in a practical manner. Modifying conditions of the trip parameters and adjusting the testing methodology were applied to relieve the response time requirement. For a further enhanced and realistic response time, revised requirements for the response time of digitalized NMS are proposed in this paper. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Human machine interface to manually drive rhombic like vehicles such as transport casks in ITER

The Cask and Plug Remote Handling System (CPRHS) and the respective Cask Transfer System (CTS) are designed to transport activated components between the reactor and the hot cell buildings of ITER during maintenance operations. In nominal operation, the CPRHS/CTS shall operate autonomously under human supervision. However, in some unexpected situations, the automatic mode must be overridden and the vehicle must be remotely guided by a human operator due to the harsh conditions of the environment. The CPRHS/CTS is a rhombic-like vehicle with two independent steerable and drivable wheels along its longitudinal axis, giving it omni-directional capabilities. During manual guidance, the human operator has to deal with four degrees of freedom, namely the orientations and speeds of two wheels. This work proposes a Human Machine Interface (HMI) to manage the degrees of freedom and to remotely guide the CPRHS/CTS in ITER taking the most advantages of rhombic like capabilities. Previous work was done to drive each wheel independently, i.e., control the orientation and speed of each wheel independently. The results have shown that the proposed solution is inefficient. The attention of the human operator becomes focused in a single wheel. In addition, the proposed solution cannot assure that the commands accomplish the physical constrains of the vehicle, resulting in slippage or even in clashes. This work proposes a solution that consists in the control of the vehicle looking at the position of its center of mass and its heading in the world frame. The solution is implemented using a rotational disk to control the vehicle heading and a common analogue joystick to control the vector speed of the center of the mass of the vehicle. The number of degrees of freedom reduces to three, i.e., two angles (vehicle heading and the orientation of the vector speed) and a scalar (the magnitude of the speed vector). This is possible using a kinematic model based on the vehicle Instantaneous Center of Rotation (ICR): a geometric approach where, at each time instant, the vehicle describes a circumference (either with a finite or infinite radius). The inverse of the kinematic model transforms the three input parameters of the center of mass into the four parameters for the wheels, preserving the omni-directional capabilities. The solution is implemented and tested using a HMI with a control disk and an analog joystick with two axis. The control disk was specially designed for this solution and implemented using a programmable micro-controller. In the first set of experiments, the HMI communicates with a computer running a simulator of the CPRHS/CTS, with the vehicle kinematics and dynamics, moving in a map of the ITER buildings. In the second set of experiments, the HMI communicates with a scaled prototype of the CPRHS running in a mock-up scenario to obtain more realistic results. Several type of tests were performed to evaluate the usability of the HMI. Different human operators without knowledge neither experience with this interface were invited to test the HMI. The operators had to drive the vehicle from an initial place to a final destination under the following conditions: with a pre-computed path to help guidance, without any path, with the information of the closest obstacles and without any help. The performance was evaluated using the time duration of the operation, the energy required to perform the described path, the risk of collision and, in case of a pre-computed path, the comparison between paths. In addition, each operator tested the HMI several times to evaluate the performance along consecutive trials. (authors)

42 ENGINEERING↗
DOE OSTI2017

Database dedicated to nuclear criticality events that occurred in fuel cycle facilities

Nuclear fuel cycle facilities can encounter events, resulting from human errors, failure of safety organization or technical failures. However, the application of the concept of defense in depth, especially the first two levels (prevention of abnormal situations and detection) requires to ensure safety in such undesirable situations. Furthermore, the periodic safety reviews of each nuclear fuel cycle facility involve the evaluation of the safety of these facilities with regard to the regulation evolution, the best safety practices, the facility evolution (modifications, aging...) and the global feedback. Part of this work is based on the analysis of the feedback and lessons learned from the events that occurred during the past period. This specific work indeed provides lessons about the operating conditions of the facility and highlights malfunctions that have occurred during operations, maintenance or restart after a shutdown. Taking into account this feedback, IRSN developed a database gathering the events which have affected the safety of French fuel cycle facilities (from conversion to waste storage, laboratories and experimental facilities, facilities under dismantling and decommissioning). Among all the recorded events, it is possible to extract in particular those impacting the criticality safety. In order to extend the analysis of these events, the IRSN criticality safety department developed a specific database dedicated to criticality risks. The 'criticality events' database is continuously updated on the basis of the declared events and their analysis by engineers from the criticality safety department. At this time, about 600 events are described and analyzed in this database. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Development of a Technical Standard for Radiological Characterization of DAW - 18062

Throughout its 70+-year history, the U.S. Department of Energy (DOE) and its predecessor agencies have performed a wide mission involving nuclear materials and weapons development, production, testing, and modernization. While the initial mission components relied heavily upon reactor physics and chemistry, the environmental programs have increasingly relied upon development of more accurate and effective measurement methodologies to determine the radiological and chemical properties of the unique waste streams that were generated during the weapons production and testing mission. Over time, the DOE laboratories and former production sites have developed a wide range of site-specific processes and procedures for conducting radiological characterization of dry active waste (DAW) that is managed as low-level radioactive waste. This has resulted in a Complex-wide variability among sites with respect to the methodologies and instrumentation used to characterize individual waste streams. The DAW-type waste was selected for development of a characterization standard since it is a simple, voluminous, and repetitive waste stream with a substantial knowledge base, established experience history at multiple site locations, and the potential for realization of programmatic benefits though standardization. The Energy Facility Contractors Group (EFCOG) Waste Management Working Group (WMWG) has proposed the development of a new DOE Technical Standard for radiological characterization of DAW. The purpose of the proposed new standard is to provide a technical resource that integrates and enhances the utilization of current site-specific methodologies and processes. A Technical Basis Document (TBD) has been prepared to support the development and utilization of the new standard by providing detailed information for use by site personnel when implementing the guidance provided in the standard. It also provides the rationale for and details concerning the recommended technical approaches for radiological characterization of representative DAW-type waste streams. The TBD will be included as an appendix to the proposed standard. Adoption of a standardized approach to DAW characterization will take advantage of multiple applications for a technology that is both mature and stable. The DAW characterization process is repetitive and lends itself to implementation at a wide range of operating locations - resulting in economies of scale, compliance with regulatory requirements, and consistent implementation of procedures which support design, operations, maintenance, and management of facilities, activities and programs. The purpose of the proposed new Technical Standard for DAW Radiological Characterization is to integrate and improve the utilization of current site-specific methodologies and processes and will include a comprehensive evaluation and comparison of current radiological characterization methodologies to determine which ones are most reliable and effective in producing relevant and valid data. In view of the long history and participation by several agencies in the regulation of nuclear programs, existing Voluntary Consensus Standards have also been reviewed for relevancy and applicability to the new proposed standard. The proposed new Standard will be developed in concert with existing scientific and regulatory guidance to produce a document that incorporates the state of current knowledge and site practices with a graded approach to applying those methodologies that are applicable and effective for characterizing a wide range of DAW-type wastes - including, but not limited to, debris, low specific activity (LSA) material, and surface contaminated objects (SCO). Under the leadership of the National Laboratory Subgroup of the EFCOG/WMWG), an EFCOG team will develop the proposed technical standard. This process will incorporate the information gathered during preparation of the TBD, which will serve as a technical resource to support the technical approach for selection of DAW characterization methodologies. Preparation of the TBD included a detailed survey of historical site practices and examination of the characterization approaches and measurement methodologies currently being applied to DAW-type wastes. Future activities include the development and submittal of a Project Justification Statement that will be submitted to the DOE Technical Standards Program for review and comment. The EFCOG/WMWG Team will respond to reviewer comments, as required. Following final concurrence by affected organizations through the DOE Technical Standards Coordination process, the EFCOG/WMWG Team will prepare the draft technical standard and respond to comments generated during RevCom review and final format review by the DOE Technical Standards Management Program. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of a Technical Standard for Radiological Characterization of DAW - 19552

Since the Manhattan Project, the U.S. Department of Energy (DOE) and its predecessor agencies have performed a wide mission involving nuclear materials and weapons development, production, testing, and modernization. While the initial mission components relied heavily upon reactor physics and chemistry, the environmental programs have increasingly relied upon development of more accurate and effective measurement methodologies to determine the radiological and chemical properties of the unique waste streams that were generated during the weapons production and testing mission. During the ongoing site cleanup program, the DOE laboratories and former production sites have developed a wide range of site-specific processes and procedures for conducting radiological characterization of dry active waste (DAW) that is managed as low-level radioactive waste. This has resulted in a Complex-wide variability among sites with respect to the methodologies and instrumentation used to characterize individual waste streams. The DAW-type waste was selected for development of a characterization standard since it is a simple, voluminous, and repetitive waste stream with a substantial knowledge base, established experience history at multiple site locations, and the potential for realization of programmatic benefits though standardization. Implementation of a consistent, standardized approach to DAW characterization will take advantage of multiple applications for a technology that is both mature and stable. The DAW characterization process is repetitive and lends itself to implementation at a wide range of operating locations - resulting in economies of scale, compliance with regulatory requirements, and consistent implementation of procedures which support design, operations, maintenance, and management of facilities, activities and environmental restoration programs. In response to a request from DOE, the Energy Facility Contractors Group (EFCOG) Waste Management Working Group (WMWG) is supporting the development of a new DOE Technical Standard for Radiological Characterization of DAW. The purpose of the proposed new standard is to provide a technical resource that integrates and enhances the utilization of current site-specific methodologies and processes. An EFCOG/WMWG technical team has prepared a Technical Basis Document (TBD) to support the development and utilization of the new standard. The TBD provides detailed reference information for use by site personnel when implementing the guidance to be provided in the standard. It also provides the rationale for and details concerning the recommended technical approaches for radiological characterization of representative DAW-type waste streams. The TBD is an essential resource and will be included as an appendix to the proposed standard. Formulation of the new standard will draw upon the information contained in the TBD in order to integrate and improve the utilization of current site-specific methodologies and processes. It will also include a comparison of current radiological characterization methodologies to determine which ones are most reliable and effective in producing relevant and valid data. In view of the long history and participation by several agencies in the regulation of nuclear programs, existing Voluntary Consensus Standards have also been identified in the TBD and will be referenced for relevancy and applicability to the new proposed standard. A Project Justification Statement (PJS) was also developed by the EFCOG/WMWG Team. Following internal Team review and approval by EFCOG, the draft PJS was submitted through Argonne National Laboratory (ANL) and the DOE Argonne Site Office Technical Standards Managers for review and approval by the DOE Argonne Site Office Manager. After resolution of comments and approval by the Manager, the PJS will be submitted to the DOE/HQ Technical Standards Program Manager (TSPM) for formal coordination. Following resolution of technical comments received during the PJS coordination process and receipt of a technical standards project number assigned by the TSPM, the Team will confirm the overall preparation schedule and develop an Internal Draft Technical Standard for coordination with the DOE Argonne Site Office. Following comment resolution, the Team will then prepare a Coordination Draft for submittal to the DOE Technical Standards RevCom review process and will respond to comments generated during this Complex-wide review. (authors)

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

How to reduce Shutdown Dose Rates in ITER Diagnostics Equatorial Ports

The ITER vacuum vessel presents 54 access ports, arranged in three levels, where port plugs are inserted hosting different types of systems: test blanket modules, antennas, vacuum pumps, neutral beam injectors and diagnostics. At the equatorial level, nine out of the eighteen access ports present diagnostics port plugs, in which the diagnostics systems are deployed into three DSMs (Diagnostics Shield Modules). The DSMs, being a major constituent in volume of the port plugs, must help to fulfill the design target of Shutdown Dose Rates (SDDR) corresponding to 100 μSv.h{sup -1} after 10{sup 6} s of cooling time for planned in-situ maintenance operations in the port interspace. Additionally, the DSMs are largely responsible for the port plug drained weight limit of 45 T, which imposes limitations to the shielding technology. Up to now, all the proposed DSM designs combine differently stainless steel, water and/or B{sub 4}C as shielding to accommodate a maximum of 1.1 m{sup 3} of DSM bulk shield, contained into a L 1845 x W 518 x H 2011 mm stainless steel box weighting 5.8 T for each DSM. These proposals show insufficient shielding properties with respect to SDDR, while none of them satisfies currently the weight limit. However, during 2016, a new DSM concept has been developed, named 'modular DSM', with significant engineering advances and changes. The DSM length has been increased by 500 mm, up to 2345 mm. The SS box has been reduced to the minimum structure required, presenting the same weight while being larger and hosting a larger volume of bulk shielding inside. Thus, the volume of bulk shielding has been enlarged up to 1.5 m{sup 3} per DSM. The shielding technology is based on B{sub 4}C bricks deployed in SS trays that are attached to the vertical plates. This results in a bulk shielding made of 65% of B{sub 4}C, 10% SS and 25% void, with effective density of 2.4 g.cm{sup -3}. In summary, the modular DSM, without a weight increase, hosts extra 0.5 m3 of bulk shielding along an additional 500 mm of length, to be used with a light but highly neutron absorber, what allows fitting the weight limit of 45 T for the port plug once drained. From now, the non-modular DSMs will be referred as 'short DSMs'. The modular DSM's advantages with respect to the short DSM versions have shown that the DSMs design can be improved in generic approach, i.e. when the penetrations of the systems are not taken into consideration. A mention apart is required when port specific penetrations are considered. Provided that they will be an important radiation transmission path, their presence could significantly diminish the improvement. To clarify this issue, in this work we have carried out a nuclear analysis of the ITER equatorial port no. 11 integration considering the modular DSM. It is a first plasma port hosting the following 6 major diagnostics systems: Neutral Particle Analyzer (NPA), H-α monitor, Low Field Side (LFS) reflectometry, Visible - Ultraviolet divertor (VUV div), Visible - Ultraviolet core survey (VUV CS) and X-rays Crystal System (XRCS). For the baseline case, a SDDR matrix has been derived assigning responsibilities to the different EP no. 11 individual systems in terms of neutron flux transmission through them, and the related activation due to that neutron flux (activation of itself, but also of the rest of systems taken one by one). Different EP no. 11 configurations have been studied in addition to the baseline case. For example, the backfill of the port plug penetrations to protect all the empty space not required for systems sight. Another issue is the determination of the role of the Co impurity content minimization from 0.05 wt.% to 0.03 wt.% recently proposed. In this summary, only some results related to the baseline are given.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Analysis of Shutdown Dose Rate in Fusion Energy Systems Using Hybrid Monte Carlo/Deterministic Techniques - Paper 138

Accurate assessments of shutdown dose rate (SDDR) are critical to support operation, maintenance, and waste disposal planning and to guide possible design changes of critical components in fusion energy systems. An SDDR calculation involves three steps: 1. a neutron transport calculation to determine the space and energy neutron flux distributions, 2. an activation calculation for computing the photon source distribution, and 3. a photon transport calculation for the estimation of the final SDDR. A companion paper describes the challenges of SDDR computations in fusion energy systems and the status of the techniques that are currently being used in SDDR analysis. Of these techniques, the rigorous 2-step (R2S) computational system entails Monte Carlo (MC) neutron and photon transport calculations coupled with a comprehensive activation step using a dedicated inventory code and library. The use of global MC variance reduction techniques was suggested for accelerating the SDDR MC neutron transport calculation. These techniques, which attempt to calculate MC tallies with nearly uniform relative uncertainties in both the low-flux space-energy regions and the high-flux space-energy regions, do not preferentially focus the MC computational efforts toward space-energy regions of high importance to the final decay dose. The ability of these approaches to accurately predict SDDR is inhibited by their prohibitive computational costs, which will be on the order of thousands of processor-years for full-scale modeling of an entire fusion plant. The companion paper describes the theoretical background of the Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method, which has been proposed to speed up SDDR MC neutron calculations. The MS-CADIS method uses the CADIS method, which has been successfully used for more than a decade in shielding calculations, but focuses on multistep shielding calculations such as SDDR analyses. The companion paper also describes a new method for calculating the uncertainties in SDDR due to uncertainties in the MC neutron calculation. This new method of uncertainty propagation uses the MS-CADIS neutron adjoint source to propagate the uncertainties in the neutron fluxes to the SDDR. This paper describes the application of the MS-CADIS method for speeding up the SDDR MC calculations and for calculating the SDDR uncertainty due to the neutron flux uncertainties. A new metric for assessing the reliability of SDDR calculations was also suggested and used in this analysis. The ITER benchmark problem was used in this analysis. The problem resembles the configuration and geometrical arrangement of an upper port plug in ITER. The analysis compared the efficiency of the MS-CADIS method to the traditional approach of using global MC variance reduction techniques for speeding up SDDR MC neutron calculation. The application of the MS-CADIS method to SDDR calculations in fusion energy systems was tested using the ITER benchmark problem. Compared to the standard FW-CADIS method, the increase in the efficiency of the SDDR neutron MC calculation due to the use of the MS-CADIS method was between 18% and 69%. The MS-CADIS method also increases the fraction of non-zero-scoring mesh tally elements in the space-energy regions of high importance to the final SDDR. Implementation of the MS-CADIS method in the SCALE and the ADVANTG code systems is currently under way. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Nuclear criticality safety standard for a fuel reprocessing plant assuming burnup credit published by the atomic energy society of Japan - 14591

In this paper an outline of the standard published in April 2015 by the Atomic Energy Society of Japan (AESJ) is presented. The standard gives basis of nuclear safety management of a spent fuel reprocessing plant that assumed burnup credit in nuclear criticality safety evaluation. Ten years ago the AESJ published Basic Items of Criticality Safety Control: 2004, AESJ-SC-F004:2004, which prescribed basic ideas, requirements and methods on nuclear criticality safety controls of facilities handling with nuclear fuel materials in general for preventing a nuclear criticality accident. However, it did not include any specific procedures for adopting burnup credit. Therefore, a new standard was envisaged as the first Standard for fuel reprocessing plants, which clarified the specific procedures to apply burnup credit to designers, operators, maintenance persons and administrators. The Standard, Procedures for Applying Burnup Credit to Criticality Safety Control of a Reprocessing Facility: 2014, AESJ-SC-F025: 2014, was developed by the AESJ after vital discussions of the Working Group on Nuclear Criticality Safety Control, which were followed by critical reviews of the Special Committee on Nuclear Fuel Cycle and those of the Standard Committee, where both committees were established under the AESJ. The followings are some features of the standard: 1. In many countries, application of burnup credit has been considered especially in transportation and storage of spent fuel. The standard applies to a reprocessing plant, especially to the Rokkasho Reprocessing Plant in Japan. The present edition of the Standard assumes the first phase of burnup credit, i.e., burnup credit considering reduction in {sup 235}U, production of {sup 239}Pu and changes in other actinides, but not considering neutron absorption effect of fission products. 2. This standard covers nuclear criticality safety management including safety design and safety control stages. It prescribes each procedure of nuclear safety management with emphasis on its necessity rather in detail. Attached to the standard are seven appendices, which give reference to it, and explanations, which provide supplemental information to the readers. Future needs of AESJ standards in the field of nuclear criticality safety will be discussed in the presentation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Updates and Lessons Learned from NuMI Beamline at Fermilab

The Neutrinos at the Main Injector (NuMI) beamline at Fermilab generates an intense muon neutrino beam for the NOvA (NuMI Off-axis $\nu_e$ Appearance) long-baseline neutrino experiment. Over the years, the NuMI beamline has been pivotal in advancing neutrino physics, providing invaluable data and insights. This proceeding paper discusses updates and the lessons learned from recent experiences during the beam operations, maintenance, and monitoring of the NuMI beamline. Key topics include the optimization of beam performance and challenges in maintaining beamline stability. The paper aims to share best practices and provide a road-map for future beamline projects, including the Long-Baseline Neutrino Facility (LBNF).

43 PARTICLE ACCELERATORS↗