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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Activation, Heating and Exposure Rates for Mo‐99 Experiments with 25‐Disk Targets

An MCNPX model of the 25-disk target assembly inside the vacuum cube inside the shielded box was prepared. This was used to calculate heating and photon and neutron fluxes throughout the model. Production rates for photonuclear reaction products were calculated using the photon fluxes and ENDF/B-VII cross sections. Measured isomer to ground state yield ratios were used where available. Where not available the new correlation between spin deficit and isomer to ground state yield ratios presented at AccApp'11 was used. The photonuclear production rates and neutron fluxes were input to CINDER2008 for transmutation calculations. A cross section update file was used to supply (n,n') reactions missing from CINDER2008 libraries. Decay photon spectra produced by CINDER2008 were then used to calculate exposure rates using the MCNPX model. Two electron beam irradiations were evaluated. The first was for a thermal test at 15 MeV with 1300 μA incident on one target end and the second was for a production test at 35 MeV with 350 μA incident on both target ends (700 μA total current on target). For the thermal test 1, 2, 3, 4, 5 and 6 h irradiation times were simulated, each followed by decay time steps out to 42 days. For the production test 6, 12, 18, 24, 30 and 36 h irradiation times were simulated followed by the same decay periods. For all simulations beam FWHMs in x and y were both assumed to be 6 mm. Simulations were run for Mo-100 enriched and natural Mo targets for both tests. It is planned that thermal test will be run for 4 h with natural target disks and production test will be run for 24 h with enriched target disks. Results for these two simulations only are presented in this report. Other results can be made available upon request. Post irradiation exposure rates were calculated at 30 cm distances from left, right, front and back of the following configurations: (1) Shielded box with everything in it (beam pipes, cooling pipes, vacuum cube, target housing weldment and target assembly), (2) Shielded box with everything in it except the target assembly, (3) Shielded box with nothing in it, (4) Target assembly taken outside of shielded box, (5) Target disks in cradle (target assembly with thermocouple weldment and flange removed), (6) Empty cradle, and (7) Target disks alone. Decay photon spectra from the CINDER2008 calculations were used as sources for the exposure rate calculations in the same model used for the flux calculations with beam on. As components were removed to simulate the seven cases considered the material compositions were changed to air and their respective sources were turned off. The MCNPX model geometry is plotted in Figure 1. The left and right detector locations for cases 1, 2 and 3 were 30 cm from the shielded box walls and 30 cm from the beam pipe openings in the left and right sides of the model (they are not in the beam line). A zoomed in plot of the target assembly alone is in Figure 2. Exposure rates for the seven cases are plotted as a function of time after irradiation in Figures 3, 4 and 5. To aid in comparison between the cases, all of these figures have been plotted using the same scale. Figures 3 and 4 are respectively the thermal and production test results for cases 1 through 6. Figure 5 includes case 7 results for both. Differences between cases 1 and 2 for both tests are not statistically significant showing that activation of components other than the target assembly, many of which are also shielding the target assembly, dominates exposure rates outside the shielded box. Case 3 shows the contribution from activation of the shield box itself. In front where shielded box wall is thickest box activation accounts for essentially all of the exposure rate outside. Differences between cases 4 and 5 are also minimal, showing that the contribution to target assembly exposure rates from the thermocouple flange and weldment are small compared to the target disks and cradle. From the numerical results the contribution is about 1%. Results for case 6, the cradle itself, are initially about half of the target assembly exposure rates. For the thermal test maximum shielded box exposure rates with or without the target assembly in it are about 1 mR/h 1 h after shutdown versus 65 mR/h for the production test. In front of the shielded box where a person would be positioned for target removal activities the calculated exposure rate is about 20% of the maximum for the production test, and after a few hours of decay is less than 1 mR/h. However, the window was not modeled. Typically lead glass is about one quarter as effective as solid lead for the same thickness and drawing indicates a window thickness about double the front wall thickness suggesting exposure rate in front of the window will likely be comparable to what it is in the back, 25 mR/h 1 h after shutdown and about 10 mR/h at 1 d.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Operational Characteristics of an Accelerator Driven Fissile Solution System

Operational characteristics represent the set of responses that a nuclear system exhibits during normal operation. Operators rely on this behavior to assess the status of the system and to predict the consequences of off-normal events. These characteristics largely refer to the relationship between power and system operating conditions. The static and dynamic behavior of a chain-reacting system, operating at sufficient power, is primarily governed by reactivity effects. The science of reactor physics has identified and evaluated a number of such effects, including Doppler broadening and shifts in the thermal neutron spectrum. Often these reactivity effects are quantified in the form of feedback coefficients that serve as coupling coefficients relating the neutron population and the physical mechanisms that drive reactivity effects, such as fissile material temperature and density changes. The operational characteristics of such nuclear systems usually manifest themselves when perturbations between system power (neutron population) and system operating conditions arise. Successful operation of such systems requires the establishment of steady equilibrium conditions. However, prior to obtaining the desired equilibrium (steady-state) conditions, an approach from zero-power (startup) must occur. This operational regime may possess certain limiting system conditions that must be maintained to achieve effective startup. Once steady-state is achieved, a key characteristic of this operational regime is the level of stability that the system possesses. Finally, a third operational regime, shutdown, may also possess limiting conditions of operation that must be maintained. This report documents the operational characteristics of a “generic” Accelerator Driven Fissile Solution (ADFS) system during the various operational regimes of startup, steady-state operation, and shutdown. Typical time-dependent behavior for each operational regime will be illustrated, and key system parameters, such as response times, will be quantified. A generalized linear systems analysis of steady-state operations will be performed to evaluate the level of stability of ADFS systems. This information should provide a basic understanding of typical ADFS system operational behavior, and facilitate the development of monitoring procedures and operator aids.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

OM Code Requirements For MOVs -- OMN-1 and Appendix III

The purpose or scope of the ASME OM Code is to establish the requirements for pre-service and in-service testing of nuclear power plant components to assess their operational readiness. For MOVs this includes those that perform a specific function in shutting down a reactor to the safe shutdown condition, maintaining the safe shutdown condition, and mitigating the consequences of an accident. This paper will present a brief history of industry and regulatory activities related to MOVs and the development of Code requirements to address weaknesses in earlier versions of the OM Code. The paper will discuss the MOV requirements contained in the 2009 version of ASME OM Code, specifically Mandatory Appendix III and OMN-1, Revision 1.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

PWR cores with silicon carbide cladding

The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

ESBWR response to an extended station blackout/loss of all AC power

U.S. federal regulations require light water cooled nuclear power plants to cope with Station Blackouts for a predetermined amount of time based on design factors for the plant. U.S. regulations define Station Blackout (SBO) as a loss of the offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency AC power system. According to U.S. regulations, typically the coping period for an SBO is 4 hours and can be as long as 16 hours for currently operating BWR plants. Being able to cope with an SBO and loss of all AC power is required by international regulators as well. The U.S. licensing basis for the ESBWR is a coping period of 72 hours for an SBO based on U.S. NRC requirements for passive safety plants. In the event of an extended SBO (viz., greater than 72 hours), the ESBWR response shows that the design is able to cope with the event for at least 7 days without AC electrical power or operator action. ESBWR is a Generation III+ reactor design with an array of passive safety systems. The ESBWR primary success path for mitigation of an SBO event is the Isolation Condenser System (ICS). The ICS is a passive, closed loop, safety system that initiates automatically on a loss of power. Upon Station Blackout or loss of all AC power, the ICS begins removing decay heat from the Reactor Pressure Vessel (RPV) by (i) condensing the steam into water in heat exchangers located in pools of water above the containment, and (ii) transferring the decay heat to the atmosphere. The condensed water is then returned by gravity to cool the reactor again. The ICS alone is capable of maintaining the ESBWR in a safe shutdown condition after an SBO for an extended period. The fuel remains covered throughout the SBO event. The ICS is able to remove decay heat from the RPV for at least 7 days and maintains the reactor in a safe shutdown condition. The water level in the RPV remains well above the top of active fuel for the duration of the SBO event. Beyond 7 days, only a few simple actions are needed to cope with the SBO for an indefinite amount of time. The operation of the ICS as the primary success path for mitigation of an SBO, allows for near immediate plant restart once power is restored. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2012

Application of the JENDL-4.0 nuclear data set for uncertainty analysis of the prototype FBR Monju

This paper deals with uncertainty analysis of the Monju reactor using JENDL-4.0 and the ERANOS code 1. In 2010 the Japan Atomic Energy Agency - JAEA - released the JENDL-4.0 nuclear data set. This new evaluation contains improved values of cross-sections and emphasizes accurate covariance matrices. Also in 2010, JAEA restarted the sodium-cooled fast reactor prototype Monju after about 15 years of shutdown. The long shutdown time resulted in a build-up of {sup 241}Am by natural decay from the initially loaded Pu. As well as improved covariance matrices, JENDL-4.0 is announced to contain improved data for minor actinides 2. The choice of Monju reactor as an application of the new evaluation seems then even more relevant. The uncertainty analysis requires the determination of sensitivity coefficients. The well-established ERANOS code was chosen because of its integrated modules that allow users to perform sensitivity and uncertainty analysis. A JENDL-4.0 cross-sections library is not available for ERANOS. Therefor a cross-sections library had to be made from the original ENDF files for the ECCO cell code (part of ERANOS). For confirmation of the newly made library, calculations of a benchmark core were performed. These calculations used the MZA and MZB benchmarks and showed consistent results with other libraries. Calculations for the Monju reactor were performed using hexagonal 3D geometry and PN transport theory. However, the ERANOS sensitivity modules cannot use the resulting fluxes, as these modules require finite differences based fluxes, obtained from RZ SN-transport or 3D diffusion calculations. The corresponding geometrical models have been made and the results verified with Monju restart experimental data 4. Uncertainty analysis was performed using the RZ model. JENDL-4.0 uncertainty analysis showed a significant reduction of the uncertainty related to the fission cross-section of Pu along with an increase of the uncertainty related to the capture cross-section of {sup 238}U compared with the previous JENDL-3.3 version. Covariance data recently added in JENDL-4.0 for {sup 241}Am appears to have a non-negligible contribution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Lessons-Learned from D and D Activities at the Five Gaseous Diffusion Buildings (K-25, K- 27, K-29, K-31 and K-33) East Tennessee Technology Park, Oak Ridge, TN - 13574

The East Tennessee Technology Park (ETTP) is the site of five former gaseous diffusion plant (GDP) process buildings that were used to enrich uranium from 1945 to 1985. The process equipment in the original two buildings (K-25 and K-27) was used for the production of highly enriched uranium (HEU), while that in the three later buildings (K-29, K-31 and K-33) produced low enriched uranium (LEU). Equipment was contaminated primarily with uranium and to a lesser extent technetium (Tc). Decommissioning of the GDP process buildings has presented several unique challenges and produced many lessons-learned. Among these is the importance of good, up-front characterization in developing the best demolition approach. Also, chemical cleaning of process gas equipment and piping (PGE) prior to shutdown should be considered to minimize the amount of hold-up material that must be removed by demolition crews. Another lesson learned is to maintain shutdown buildings in a dry state to minimize structural degradation which can significantly complicate characterization, deactivation and demolition efforts. Perhaps the most important lesson learned is that decommissioning GDP process buildings is first and foremost a waste logistics challenge. Innovative solutions are required to effectively manage the sheer volume of waste generated from decontamination and demolition (D and D) of these enormous facilities. Finally, close coordination with Security is mandatory to effectively manage Special Nuclear Material (SNM) and classified equipment issues. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Potential Cost Implications of an Interim Storage Facility for Commercial SNF+ - 17132

The question of whether or not consolidated interim storage of commercial spent nuclear fuel (SNF) should be part of the federal waste management system as an intermediate step before permanent disposal has been debated for more than four decades. This paper summarizes an evaluation of the cost implications of incorporating a consolidated interim storage facility (ISF) into the waste management system (WMS). In this study, the order-of-magnitude estimates of total system costs were calculated and tabulated. The analyzed scenarios involve shipment of SNF from reactors in dual-purpose storage and transportation canisters (DPCs) currently being used by utilities for dry storage at reactor sites. A number of pertinent conclusions can be drawn from this evaluation, including: - Delay in repository availability increases total system costs. Any delay in opening a repository increases total system costs, regardless of whether the system has an ISF or not. This is due to the increased cost associated with an extended duration for storage, whether at multiple independent spent fuel storage installations (ISFSIs) or at an ISF, until waste can be disposed of.- There is a potential total system lifecycle cost avoidance in realistic scenarios with an ISF when compared to scenarios with no ISF. However, most of the cost avoidance occurs several decades after the ISF is opened. The total WMS cost differential over the long term is mainly attributed to the reduced operational costs of storing the fuel in a consolidated facility rather than at individual reactor sites. - Earlier establishment of an ISF allows for more avoidance of post-shutdown at-reactor storage costs for any repository opening date. An ISF allows earlier acceptance of fuel from reactors, which reduces at-reactor costs from a total system perspective. - Transportation costs have little impact on a WMS with or without an ISF. These impacts range from 3-11% of the total cost in all scenarios. Therefore, transporting the fuel twice does not appear to be a significant cost concern driver relative to other system costs. In conclusion, designing, licensing, and constructing an ISF will require a large near-term financial investment. However, reduced annual operating and maintenance costs for an integrated waste management system with an ISF, compared to a system without an ISF, could eventually result in system-wide (integrated) cost avoidance due to reductions in at-reactor storage costs in the long term. The largest cost avoidances would not occur for several decades, so assumptions about inflation, escalation, and discount rates will have a significant effect on potential economic impacts of an ISF. It should be noted that other benefits associated with the investment of an ISF-such as earlier removal of SNF from shutdown reactor sites, earlier prospects for site re-utilization, and additional overall integrated waste management system flexibility-were not addressed in this study. (authors)

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

The decommissioning imperative - motivating progress in us used fuel management - 15185

When the Yucca Mountain project was terminated in 2010 there was approximately 3500 metric tons of used fuel stored at Independent Spent Fuel Storage Installations (ISFSIs) at which there was no operating reactor (10 shutdown plant sites and the GE Morris facility). By 2015, four additional sites with five reactors will have shut down, causing the amount of used fuel 'stranded' at sites where there is no operating reactor to nearly double to approximately 6900 metric tons which will be stored at a total of 15 sites. At each of these sites, once the reactors are decommissioned, the only thing preventing the land from being returned to communities for other uses is the remaining presence of the used fuel. The cost of maintaining 15 stand-alone ISFSIs is born by electric ratepayers and taxpayers. Going forward, the costs of maintaining stand-alone ISFSIs is certain to grow even if the number of stand-alone ISFSIs does not grow beyond the already announced shutdown of the Oyster Creek reactor in 2019 (which would increase the amount of stranded used fuel by another 10% and create a 16. stand-alone ISFSI). One reason for this cost escalation will be the need for ISFSI owners to renew storage licenses, and implement aging management programs to support these renewals, due to the continued delay in the federal program for removing the used fuel. Storage licenses at 10 of 16 stand-alone ISFSIs will expire by 2021, which is the earliest projected date at which the Department of Energy (DOE) has indicated it could have a pilot interim storage facility available - and nearly 3 decades before DOE projects a repository will be available. The burden and cost of continuing to renew storage licenses - and manage aging storage facilities - at 16 stand-alone sites instead of moving stranded used fuel to a single consolidated storage location or repository will likely be significant. Industry resources are increasingly being channeled into dry storage aging management activities as public and regulator interest in potential long-term degradation mechanism continues to grow. Whether the ultimate destination for used fuel is Yucca Mountain or some other location, the current stalemate preventing progress in the federal program to move it must be broken - and the increasing prominence of the decommissioned reactor quandary may very well be a pivotal factor in breaking that stalemate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Immediate Dismantling of a Large Fleet of LWR NPPs: Consequences for Spent Fuel and Waste Management - 16027

The experience feedback shows that the complete dismantling of one Light Water Reactor (LWR) unit is an activity now globally well under control (reactor pressure vessel and its internals removal and cutting included). However some specific industrial difficulties may arise in view of the dismantling of numerous LWR units simultaneously. Indeed more than 300 LWR units are currently operated worldwide, which have been commissioned these last 50 years. So in some countries, many units may be permanently shut down over a period of few years in the next decades. Such a situation addresses the issue of the overall management of large quantities of spent nuclear fuel (SNF) and decommissioning waste for the concerned units. For example, in France, the legislative and regulatory framework for the nuclear facilities favors their dismantling 'as soon as possible' after their permanent shutdown which implies as well to limit the duration of the transition period from operation to decommissioning. Furthermore 58 LWR units have been commissioned between 1977 and 1999 in France - on average more than 2 units per year. In this context, the operator (Electricite de France - EDF) plans to remove quickly the SNF then to perform dismantling actions immediately after the permanent shutdown of the LWR units. One issue is to remove the SNF from all the relevant units, even if this removal is simultaneous in many units (permanently shut down or still under operation). Similar issue has to be taken into account regarding the management of the radioactive waste (RW) produced by the dismantling and clean-up actions, notably the RW that cannot be disposed of in a near surface repository. One method followed by the French technical support organization (Institut de Radioprotection et de Surete Nucleaire - IRSN) to analyze these issues is the use of estimates of flows of SNF and RW. These estimates are based notably on a phasing-scenario and a planning template defined for the decommissioning of the units of one nuclear power plant (NPP) and coupled to an overall schedule for phase out all the units of the fleet. The estimates relative to the next decades can be compared to the current experience feedback of flows of SNF and RW for units under operation, in order to identify risks when facing decommissioning. The risks highlighted are driven by key parameters (as duration of the main dismantling actions) of the estimates which can be adapted to minimize their impact. On this basis, it is possible to identify the key-factors to dismantle each unit of NPPs and phase out the fleet regarding SNF and RW management. It is noteworthy that this work needs to be done in any case upstream the studies and the implementation of dismantling actions. Finally it can be underlined that another issue is the human resources (staff, skills and knowledge) necessary to perform all the decommissioning actions, but this aspect is not addressed in the present paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

A WM Facility for the Processing of Radioactive Materials Arising from the Decommissioning Activities of the Belgian Nuclear Power Plant of Doel - 16042

Doel Nuclear Power Plant is located in the northern part of Belgium in the harbour of Antwerp on the Scheldt River. On this 80-hectare site Electrabel, the Belgian utility, operates four PWRs with a total generation capacity of 2911 MWe. Those nuclear power units were commissioned in 1975 (Doel 1 and 2 - twin units), in 1982 (Doel 3) and in 1985 (Doel 4). In light of the Belgian governments decision to definitively shutdown the oldest units of the site (Doel 1 and 2 - 433 MWe each) in 2015 at the end of their 40-year design lifetime, Electrabel launched a D and D program in partnership with Tractebel Engineering in July 2012. During the D and D preparation phase prior to the final shutdown, studies performed by Tractebel Engineering showed the added benefit of a purpose-built Waste Management Facility (WMF) specifically equipped to process contaminated and activated components generated by the in-situ dismantling activities. The WMF is a key infrastructure contributing to a large extent to the Waste Minimization Policy within the global D and D program. The present paper aims at presenting its strategic, logistic, infra-structural, and design principles. At the end of 2014, the Belgian government and Electrabel entered into discussions to envisage a 10-year lifetime extension of Doel 1 and 2 units. Therefore, Electrabel decided to put the D and D program on hold. Nevertheless, the preparatory studies for the future D and D activities on Doel site are set on solid assumptions and pave the way towards a safe, cost-effective and waste minimizing decommissioning process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Transitioning From Operations to Decommissioning - 16049

With the recent unplanned shutdown of a number of U.S. Nuclear Power Plants, a benchmarking effort was conducted in late 2104 to capture the lessons learned from these recently shutdown sites. The specific plants benchmarked for this effort included San Onofre Nuclear Generating Station (SONGS), Kewaunee, Zion, and Crystal River Nuclear Power Stations. The benchmark results found for the most part, site personnel had little or no decommissioning experience and in fact, were not aware of the challenges facing a plant in transitioning from operations to decommissioning. This lack of experience and knowledge resulted in a steep learning curve, a loss of productivity, and significant challenges by plant personnel prepared the site for decommissioning. The results of the benchmark identified a number of areas and challenges facing an operating plant during this period. All sites visited identified four key areas where significant cost savings and planning effectiveness could have been realized if they had been better prepared. The areas included transition planning, staffing, stakeholder involvement, and cost estimating. It was recognized if these areas are aggressively managed early in the process, it will have a significant positive impact on the effectiveness and cost of decommissioning and can better set the stage for a successful project. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Overview of the U.S. Department of Energy, Office of Nuclear Energy's Nuclear Fuel Storage and Transportation Planning Project

The U.S. Department Energy, Office of Nuclear Energy (DOE-NE) is laying the groundwork for implementing interim storage, including associated transportation, per the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High- Level Radioactive Waste, and to develop a foundation for a new nuclear waste management organization. The organization within DOE-NE that is responsible for this planning is the Office of Nuclear Fuels Storage and Transportation (NFST). The technical work to support this planning effort is being conducted by DOE's national laboratories. DOE's overarching goal is to develop options for decision-makers on the design of an integrated waste management system. The goal is to develop an integrated plan to (1) implement interim storage; (2) improve the overall integration of storage as a planned part of the waste management system; (3) prepare for the large-scale transportation of SNF and high-level waste (HLW), with an initial focus on removing SNF from the shutdown reactor sites; and (4) develop the foundational information, resources, and capabilities needed to support future implementation decisions and actions. Near-term objectives are to (1) develop and maintain an integrated plan to accomplish the Administration's Strategy goals, (2) improve integration of storage as a planned part of the waste management system, (including evaluating standardization of dry cask storage systems), (3) develop and evaluate design options for an integrated waste management system, (4) develop and apply systems analyses to provide quantitative estimates of system impacts of utility actions' impacts on the system and inform future decisions, (5) prepare for the large-scale transportation of SNF and HLW, (with an initial focus on removing SNF from the shutdown reactor sites), and (6) establish and maintain a unified, integrated SNF database and analysis system to characterize the input to the waste management system. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Alternate Packaging Options for Remote-Handled Transuranic Waste Disposal at WIPP - 19093

Sandia National Laboratories/New Mexico (Sandia) began a remote-handled (RH) transuranic (TRU) repackaging campaign two weeks before the Waste Isolation Pilot Plant (WIPP) shutdown in February 2014. Since a contract with Nuclear Waste Partnerships (NWP), the WIPP Management and Operation (M and O) contractor, was already in place for certified visual examination (VE) and dose-to-curie (DTC) support, repackaging continued and was completed in October 2015. Twenty-five 55-gallon drums of RH TRU waste were packaged, awaiting certification and shipment to WIPP. However, in February 2017 at the National TRU Program (NTP) Corporate Board meeting, Sandia learned that RH TRU waste normally transported in removable lid canisters (RLCs) and 72-B casks was not being accepted for disposal when WIPP re-opened after the shutdown. One option, shielded container assemblies (SCAs), was available and approved for RH TRU waste disposal at the WIPP. Sandia made the decision to pursue the use of SCAs. The use of SCAs allow RH TRU waste to be shipped as contact-handled (CH) in HalfPACTs, emplaced on the floor of the WIPP instead of in boreholes in the walls, if the dose rate is less than 200 millirem/hour on contact. The RH TRU waste remains on the WIPP inventory as RH. (authors)

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

Migration to a fully-integrated control room - 349

The nuclear industry is faced with a pressing need to reduce operating costs in order to remain competitive with other forms of electric generation. While current industry initiatives are focused on incremental gains by streamlining work processes and reducing worker burdens, very few, if any, involve significant reductions in the base workload of a nuclear plant. Migration to a fully-integrated control room affords just such an opportunity. This report describes an NPP digital instrumentation and control (I and C) system design, along with its corresponding human systems interface (HSI) design, which achieves high plant availability, while selectively applying digital equipment redundancy and integrating traditional I and C system partitions, to minimize the digital equipment needed, and thereby minimize initial installation costs and ongoing operations and maintenance (O and M) costs. An additional key component of digital equipment minimization is optimization of the digital interface to conventional plant sensors and controlled plant devices (e.g., pumps, valves), which requires no disturbance to plant cables and results in eliminating thousands of electro-mechanical relay control components. The design is rooted in the highly integrated digital technology applied to the most recent advanced light water reactors under United States (U.S.) Nuclear Regulatory Commission (NRC) design certification review and approval. While this design employs technology from new plants, that technology has been adapted to the modernization of current operating NPPs. Hence, the highly integrated digital I and C and HSI design is referred to as the compact control room digital modernization (CDM). Along with minimal digital equipment, the corresponding HSI design employs compact operator workstations with visual display units (VDU) and soft controls for both the main control room (MCR) and remote shutdown room (RSR), to immediately improve human performance and ultimately facilitate operating staff reduction. Some advanced light water reactors have demonstrated this staff reduction through full scope simulator testing of normal and abnormal plant conditions, with U.S. licensed operating crews from currently-operating U.S. plants. Digital equipment minimization, compact operator workstations, and optimized plant component interfaces that require no disturbance to field cables, also facilitate implementation of the CDM over multiple phases, most of which can be conducted during normal refueling shutdown windows. This phased implementation method includes a novel transition of the I and C, MCR and RSR with an operations centered focus. This paper presents the CDM design, along with qualitative cost/benefit comparisons to currently operating plants with conventional analog I and C and HSI technology; this generation of I and C/HSI is referred to as a conventional control room (CCR). A CCR also includes NPPs that have been upgraded with digital systems, but retain the current partitions between plant I and C systems (i.e., no I and C integration) and retain the current plant system to control board configuration of current NPP control rooms. Even though these comparisons are qualitative, a compelling case is presented for the CDM. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The Effect of Localized Pin Flux on Short-Term Gamma Intensities of Spent Fuel Assemblies

Nuclear reactor operation for electricity generation involves irradiation of a fuel assembly for an extended period of time until a certain burnup is achieved. To harness the full nuclear energy potential and flatten the flux around the core, fuel assemblies are shuffled to one or more different locations within the core. Post-irradiation practice typically involves storing the spent fuel assembly in a spent fuel pool. While in the spent fuel pool, aspects of the fuel assembly continue to evolve, particularly the isotopic concentration as nuclides decay. Numerous nondestructive assay (NDA) safeguards techniques exist and have been proposed for performing verification on spent fuel assemblies to ensure they have not been tampered with. Adigun et al. recently proposed measuring both the neutron and gamma signatures from spent fuel from discharge to final disposal, however, the research in only investigated the neutron part of this approach. In the conclusion, the authors caveated the need for the gamma results to make more certain conclusions about a spent fuel assembly, as a combination of the neutron and gamma NDA measurements will limit the range of possible initial enrichment and burnup combinations. This paper focuses on the production of the gamma source. It is important to understand the driving mechanisms behind the emissions signatures at shutdown and after long cooling time periods, to better determine the correlation between the measured passive gamma emission spectra and the gamma emissions per second of the assembly. The objective is to demonstrate how the effect of localized flux within a reactor affects the gamma emissions per second (or gamma intensity) present in an assembly at shutdown. This study examined the photons emitted from the nuclides present within the fuel assembly, and not the photons escaping from the fuel assembly. The latter point of transporting the photons present within the spent fuel assemblies to an appropriate detector will be addressed in a future paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗