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Uncertainty Quantification and Management for Multi-scale Nuclear Materials Modeling

Understanding and improving microstructural mechanical stability in metals and alloys is central to the development of high strength and high ductility materials for cladding and cores structures in advanced fast reactors. Design and enhancement of radiation-induced damage tolerant alloys are facilitated by better understanding the connection of various unit processes to collective responses in a multiscale model chain, including: dislocation nucleation, absorption and desorption at interfaces; vacancy production, radiation-induced segregation of Cr and Ni at defect clusters (point defect sinks) in BCC Fe-Cr ferritic/martensitic steels; investigation of interaction of interstitials and vacancies with impurities (V, Nb, Ta, Mo, W, Al, Si, P, S); time evolution of swelling (cluster growth) phenomena of irradiated materials; and energetics and kinetics of dislocation bypass of defects formed by interstitial clustering and formation of prismatic loops, informing statistical models of continuum character with regard to processes of dislocation glide, vacancy agglomeration and swelling, climb and cross slip.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Fundamental Understanding of Creep-Fatigue Interactions in 9Cr-1MoV Steel Welds. Final Report

Ferritic-martensitic steels constitute a vital class of materials for current and future generation of nuclear reactors as well as ultra supercritical coal fired power plants due to their superior creep resistant properties. One particular ferritic-martensitic steel, 9Cr-1MoV (Grade 91) is one of five materials approved for high temperature use in nuclear service by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC). Grade 91 is also a candidate material for proposed fission and fusion reactors such as the sodium cooled fast reactor (SFR). In service, Grade 91 components can be subjected to elevated temperatures and cyclic stresses, leading to the accumulation of creep-fatigue (CF) damage. While the monotonic creep properties of Grade 91 steels have been studied extensively, there is less knowledge concerning the mechanisms governing CF damage of this steel in the creep-dominated failure region. Moreover, the creep failure is especially pronounced in welded Grade 91 components, which are commonly joined using conventional arc welding techniques such as flux-cored arc welding (FCAW). In a Grade 91 weld, the creep failure typically occurs in the outer region of the heat affected zone (HAZ), specifically in the intercritical or fine-grained heat-affected zone (ICHAZ or FGHAZ), a phenomenon known as type IV failure. Currently, there exists a limited knowledge of CF damage mechanisms in Grade 91 welds. Improving the mechanistic understanding of the CF damage and failure of Grade 91 steel and its welds with inhomogeneous microstructure is crucial for the safety and design of key structural components in advanced, high-temperature reactors. The overall objectives of the project were to advance the state of knowledge and understanding of CF damage in both Grade 91 base metal and its welds, especially under loading conditions where creep is the dominant damage mechanism. Uncovering the mechanisms governing the creep-fatigue deformation and damage in Grade 91 base metal in creep-dominated region as well as type IV failure in Grade 91 welds contributes to a new knowledge that may lead to more accurate life prediction models than those currently in use. Moreover, the superior CF properties obtained by low heat input welding has a high potential to develop practical solutions to the premature failure of Grade 91 weld.

36 MATERIALS SCIENCE↗

One-Step Ahead Prediction of Thermal Mixing Tee Sensors with Long Short Term Memory (LSTM) Neural Networks

High-temperature advanced reactors under development, such as sodium fast reactors (SFR) and molten salt cooled reactors (MSCR), are expected to offer lower levelized cost of energy (LCOE) compared to existing light water reactor (LWR’s). In the existing light water reactors (LWR’s), operation and maintenance (O&M) expenses constitute the largest fraction of the total operating cost. Some of the O&M costs are related maintenance of sensors which can fail due to exposure to harsh environment in a reactor. The O&M costs of Advanced Reactor (AR)’s are expected to constitute a significant fraction of the total cost as well, because of high temperature and radiation level in AR are likely to cause material fatigue and premature failure of sensors and components. The O&M costs in AR’s could be reduced through integration of advanced informatics of performance-related sensors into a digital twin designed for reactor monitoring. For example, machine learning (ML) could be employed for real-time validation and correction of performance-related sensors, and reducing the number of performance-related physical sensor units through virtual sensing. As part of the effort, we investigate real-time validation of thermal hydraulic sensors through one-step ahead forecasting of sensor values using long short-term memory (LSTM) recurrent neural networks (RNN). The sensors are installed in a flow loop containing a thermal mixing tee, which is a common experimental model to study thermal fatigue in a thermal hydraulic loop. In addition, nonlinear transients generated in a thermal mixing tee constitute a good challenge data set for training and validation of ML algorithms. Sensors in this study include thermocouples, flow meters, and optical fibers for distributed temperature sensing. In one experiment, measurement data sets were obtained for a loop was filled with water, and in another experiment, measurements were performed on a loop filled with liquid metal Galinstan. We have also conducted preliminary investigation of one-step ahead prediction of fiber optics-based distributed temperature sensing with LSTM networks. In predicting fiber-based temperature measurements, we treated each gauge pitch of the fiber as an independent sensor. Accuracy of one-step ahead forecasting was estimated by calculating root mean square error (RMSE) for the test segment of time series of each sensor. RMSE’s for temperature sensors in water loop were, for the most part, lower than for the same sensors in Galinstan loop. The RMSE’s for flow meters were similar for both loops. The RMSE’s for distributed temperature measured with the fiber optic sensor were similar to those of the point sensors. Results of this study demonstrated the capability of LSTM one-step ahead forecasting with RMSE comparable to uncertainty in sensor measurements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Use Cases and Model Development of Thermal Storage Coupling for Advanced Nuclear Reactors

This report discusses the different options for coupling thermal energy storage (TES) systems to advanced nuclear power plants (A-NPPs) in order to enable flexible and hybrid plant operation. An advanced light-water reactor (ALWR) and a high-temperature gas-cooled reactor (HTGR) were selected as the initial use cases for demonstrating a thermally balanced energy storage coupling design for thermal power extraction. Cost functions for the A-LWR were derived from the fully balanced models that were developed based on three different coupling options with three different thermal energy bypass ratios. For the next steps, cost functions for the HTGR will also be derived, and additional nuclear reactors (e.g., a liquid-cooled fast reactor [LFR] or molten-salt reactor [MSR]) will be evaluated for coupling with TES in similar fashion, including the evaluation of their steady-state condition models and cost functions. The models presented herein showcase several design considerations, focusing on optimal deployment methodologies for achieving steady-state operation with minimum disruption to the nuclear power generation cycle. This report presents the results of steady state models developed using Aspen HYSYS®, wherein the thermal energy bypass for an NPP-TES coupling was varied up to 50%. The various components were sized using Aspen Process Economic Analyzer (APEA) and Aspen Exchanger Design and Rating (EDR), when applicable. Cost functions from these models were developed using the latest publicly available data obtained from APEA V11. The current steady-state models and cost functions provide a baseline for additional work focusing on dynamic operation and process optimization by using Idaho National Laboratory (INL)’s Framework for Optimization of Resources and Economics (FORCE) tools to evaluate the technoeconomic viability and transient operations of TES-coupled A-NPPs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Analyses of Model-Based Predictive Control for a S-CO 2 Brayton Cycle Power Converter

The supercritical carbon dioxide (S-CO 2 ) Brayton cycle is a promising alternative to steam cycles as an energy converter for advanced nuclear reactors as well as fossil-fired and concentrated solar power plants. For advanced reactors, such as Sodium-cooled Fast Reactors (SFRs), the S-CO 2 cycle offers the benefits of high cycle efficiency, very compact turbomachinery, and elimination of sodium-water reactions, all of which may enable improved economics. The efficiency gains calculated for the S-CO 2 cycle are particularly significant if advantage can be taken of the sharp changes in the carbon dioxide thermo-physical properties near the critical point. On the other hand, the same property variations give rise to challenges in the cycle design, analysis, and control. The onset of two-phase flow (where the minimum cycle temperature and pressure drop drop below the critical value) can degrade compressor performance and potentially cause damage to the compressor blades. In previous work, the feasibility of the Model-based Predictive Control (MPC) approach was assessed for improving regulation of the precooler outlet temperature. A linear reduced-order model for the precooler was generated and for this simplified standalone system operating with an MPC controller the response of outlet temperature was evaluated through simulations. These evaluations included the presence of temperature measurement time delay representing a degraded sensor. It was demonstrated that the MPC controller maintained stable precooler temperature for those delay cases where a proportional-integral controller became unstable. In the present work the simulations have been extended to evaluate the MPC controller operating under more realistic conditions, namely in the split-flow recompression cycle non-linear plant. This compares with the previous standalone reduced-order linear system described above. In the present work the MPC-based control scheme was implemented in a GPASS 1-D systems code representation of the split-flow recompression cycle. The performance of the controller was again investigated for degraded sensor conditions and compared to that of a proportional-integral controller. The results were qualitatively similar to the previous results with the MPC controller and again providing superior stability characteristics. This work provides a sound starting point for the development of model-based control systems. In particular, the MPC control approach applied to Multiple Input Multiple Output systems can be used to synchronize the operation of the different system components to improve control system dynamic performance.

20 FOSSIL-FUELED POWER PLANTS↗

Practical Multi-Cycle Refueling Simulation of Stable Salt Reactors Based on REBUS Depletion Calculation

A Python software tool ReloadPy has been developed to assistant in formulating online refueling strategies for a family of stable salt reactors. ReloadPy wraps over the legacy fast reactor analysis codes REBUS, DIF3D, and PERSENT. These codes are used to perform single cycle depletion analysis and calculate reactivity worth and power density. ReloadPy determines a refueling assembly position to achieve the user-specified primary objective while satisfying constraints on various performance parameters such as the maximum power density, the maximum reactivity worth of reloaded assembly, and the minimum cycle length. Various capabilities of prescribing refueling sequence or a range of candidate refueling positions and restarting a previous simulation with altered options are implemented to reduce the required user’s efforts to investigate different refueling strategies. This report presents the computational models and methods and also serves as a user guide. Long-term simulation capabilities with different refueling strategies are demonstrated by a series of sample SSR refueling calculations with two-dimensional surrogate core models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

SCALE 6.3 Modeling Strategies for Reactivity, Nuclide Inventory, and Decay Heat of Non-LWRs

To assess modeling and simulation capabilities for thermal hydraulics, accident progression, source term, and consequence analysis for non–light-water reactor (LWR) technologies, the US Nuclear Regulatory Commission (NRC) initiated a collaborative project between Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL) in FY20, which is detailed in “Volume 3: Computer Code Development Plans for Severe Accident Progression, Source Term, and Consequence Analysis.” This project demonstrated the capabilities of the MELCOR and SCALE codes to calculate accident scenarios during operation of relevant non-LWRs. The following five non-LWR concepts were selected for capability demonstration based on recently renewed industry interest in the United States to develop and deploy such reactor technologies: pebble-bed high temperature gas-cooled reactors (HTGRs), pebble-bed fluoride salt-cooled reactors (FHRs), molten salt–fueled reactors (MSRs), heat pipe reactors (HPRs), and sodium-cooled fast reactors (SFRs).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Design of a System Interface for Flexible Thermal Power Extraction from Two Advanced Reactors

This report presents the design of system interfaces enabling flexible thermal power extraction from two advanced nuclear reactor concepts: a sodium-cooled fast reactor (SFR) operating on a Rankine cycle, and a gas-cooled reactor (GCR) utilizing a Brayton cycle. The study addresses the dual challenge of ensuring reliable base-load electricity generation while meeting the variable high-temperature steam demands of industrial processes (IPs).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Synthesis Roadmap for Actinide Chloride Salts

Molten salt reactors (MSRs) are among the main advanced nuclear reactor types at the forefront of development by industry, with the support of the US government, for the next fleet of nuclear reactors to support the demand for energy in the coming decades. MSRs are highly unique because they are cooled and typically also fueled by molten salt. This quality brings about safety benefits such as low-pressure operation and self-stabilization of the neutron flux, operational benefits such as high-temperature operation and the ability for online refueling, and fuel and waste management benefits thanks to the flexibility of post-processing of molten salts and reprocessing options that involve removal of fission products and actinide separation. In fact, domestic deployment of molten salt (or molten salt–cooled) reactors is an approaching reality: a handful of molten salt reactor developers are planning to operate demonstration-scale reactors, as a step toward commercial-scale power reactors, within the decade. For example, Natura Resources received a construction permit in September 2024 for the deployment of MSR-1, which is a graphite-moderated thermal spectrum reactor, at Abilene Christian University. Also, TerraPower, in a collaborative effort with Southern Company and Idaho National Laboratory (INL), received approval from DOE in 2023 to proceed with the construction of the Molten Chloride Reactor Experiment (MCRE), a homogeneous chloride fast reactor at INL, and has begun assembly of system components. There are several other examples of developers at different stages of development of their own unique MSR designs (Jenet et al., 2025). As developers are in the process of obtaining approvals for their designs, deploying demonstration-scale reactors, and planning for their commercial-scale power reactors, there is a critical supply chain need for the synthesis of fuel for these reactors that must be addressed. The challenge generally is three-fold: (1) the quantity of fueled salt needed for these reactors is extraordinarily high (>100s of kilograms), but demonstrations of scaled-up techniques for fueled salt synthesis are significantly more limited than demonstrations of lab-scale syntheses; (2) each developer has a unique reactor design, which means different actinide halide elements in different carrier salts must be synthesized; and (3) there is a need for particularly high-purity salt so as to ensure the long-term operability of these reactors with minimal degradation to salt-wetted components, which necessitates synthesis techniques with high levels of quality control, repeatability, and well-characterized precursor and reactant materials. It is crucial that this supply chain challenge be addressed by demonstrating synthesis techniques that are scalable, de-risked, and well-documented so that these technologies may be adopted and utilized by industry to support the fueling needs of MSRs that are to come online within the next 10 years. With this supply chain challenge clearly defined for the developing MSR industry, it is important to consider that addressing such a challenge is oftentimes complex and context-dependent. There is not necessarily a single synthesis technique that can be scaled up and adopted to address the needs of all MSR developers; the synthesis approach that may be viable for producing a desired fuel salt will be entirely dependent on the exact salt composition needed, the quantity needed, purity needed, the refueling and waste plans, and the availability of a carrier salt for the fuel. Therefore, it is important to consider more broadly what synthesis techniques are available and have been demonstrated to understand the benefits, challenges, and general nuances that should be considered when evaluating a particular route for efficient production of high-purity fuel salts at scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

First-Principles Cost Analysis of Advanced High-Temperature Nuclear Plants

Due to the vast number of recent nuclear reactor innovations, particularly those pertaining to generation IV reactor types like high-temperature gas cooled reactors (HTGRs) and sodium-cooled fast reactors (SFRs), and the newer deployment strategies envisioned, such as use of small modular reactors (SMRs) or even microreactors, reliable, detailed, and complete costs of these nuclear innovations are needed in wide availability. The types of models that generally achieve these objectives are those incorporating the fundamental nature of the real-world systems they aspire to predict, such as first-principles models. Furthermore, first principles models typically offer predictiveness that is not attained by most other types of individual-models. However, detailed, first-principles cost modeling of nuclear reactors and entire nuclear plants is relatively limited. To address this limitation in the availability of detailed, predictive models based on fundamentals, we recently developed a range of cost models, mostly based on first-principles methodologies, to project full lifecycle costs (LCCs) of nuclear power plants (NPPs) based on multiple parallel SM-HTG-pebble bed reactors (PBRs) and SM-SFRs.

Prosser, Jacob H. [Strategic Analysis, Inc., Arlin↗

Nuclear Reactor Heat Extraction for Synthetic Fuel Plants

This report looks at the viability and best approach when producing synthetic fuels using heat and power supplied by advanced nuclear energy systems. The report looks at a low temperature integration pathway with four types of advanced reactors: a pressurized water reactor (PWR), an advanced light water reactor (A-LWR), a sodium fast reactor (SFR), and a high-temperature gas-cooled reactor (HTGR). A failure modes and effects analysis (FMEA) of the coupling system between the nuclear plant and the synthetic fuel systems is performed with the goal of identifying the reliability of such a thermal delivery system. Furthermore, a high temperature pathway is investigated for synfuel coupling to determine if this is more efficient and more cost effective as a coupling approach.

10 - SYNTHETIC FUELS↗

Evaluating Nuclear Forensic Signatures for Advanced Reactor Deployment: A Research Priority Assessment

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Transient Load Following and Control Analysis of Advanced S-CO2 Power Conversion with Dry Air Cooling

Supercritical carbon dioxide (S-CO2) Brayton cycles are under development as advanced energy converters for advanced nuclear reactors, especially the Sodium-Cooled Fast Reactor (SFR). The use of dry air cooling for direct heat rejection to the atmosphere ultimate heat sink is increasingly becoming a requirement in many regions due to restrictions on water use. The transient load following and control behavior of an SFR with an S-CO2 cycle power converter utilizing dry air cooling have been investigated. With extension and adjustment of the previously existing control strategy for direct water cooling, S-CO2 cycle power converters can also be used for load following operation in regions where dry air cooling is a requirement

Moisseytsev, Anton↗

Development/Modernization of an Advanced Non-Light Water Reactor Probabilistic Risk Assessment

In 2015, GE Hitachi Nuclear Energy (GEH) teamed with Argonne National Laboratory (Argonne) to perform Research and Development (R&D) of next-generation Probabilistic Risk Assessment (PRA) methodologies for the modernization of an advanced non-Light Water Reactor (non-LWR) PRA. This effort built upon a PRA developed in the early 1990s for GEH’s Power Reactor Inherently Safe Module (PRISM) Sodium Fast Reactor (SFR). The work had four main tasks: internal events development modeling the risk from the reactor for hazards occurring at-power internal to the plant; an all hazards scoping review to analyze the risk at a high level from external hazards such as earthquakes and high winds; an all modes scoping review to understand the risk at a high level from operating modes other than at-power; and risk insights to integrate the results from each of the three phases above. To achieve these objectives, GEH and Argonne used and adapted proven PRA methodologies and techniques to build a modern non-LWR all hazards/all modes PRA. The teams also advanced non-LWR PRA methodologies, which is an important outcome from this work. This report summarizes the project outcomes in two major phases. The first phase presents the methodologies developed for non-LWR PRAs. The methodologies are grouped by scope, from Internal Events At-Power (IEAP) to hazards analysis to modes analysis. The second phase presents details of the PRISM PRA model which was developed as a validation of the non-LWR methodologies. The PRISM PRA was performed in detail for IEAP, and at a broader level for hazards and modes. In addition to contributing methodologies, this project developed risk insights applicable to non-LWR PRA, including focus-areas for future R&D, and conclusions about the PRISM design.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Comparison of JSFR design with EDF requirements for future SFR

A comparison of Japan sodium-cooled fast reactor (JSFR) design with future French SFR concept has been done based on the requirement of EDF, the investor-operator of future French SFR, and the French safety baseline, under the framework of EDF-JAEA bilateral agreement of research and development cooperation on future SFR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Benchmark studies of thermal jet mixing in SFRs using a two-jet model

To guide the modeling, simulations and design of Sodium Fast Reactors (SFRs), we explore and compare the predictive capabilities of two numerical solvers COMSOL and OpenFOAM in the thermal jet mixing of two buoyant jets typical of the outlet flow from a SFR tube bundle. This process will help optimize on-going experimental efforts at obtaining high resolution data for V and V of CFD codes as anticipated in next generation nuclear systems. Using the k-{epsilon} turbulence models of both codes as reference, their ability to simulate the turbulence behavior in similar environments was first validated for single jet experimental data reported in literature. This study investigates the thermal mixing of two parallel jets having a temperature difference (hot-to-cold) {Delta}T{sub hc}= 5 deg. C, 10 deg. C and velocity ratios U{sub c}/U{sub h} = 0.5, 1. Results of the computed turbulent quantities due to convective mixing and the variations in flow field along the axial position are presented. In addition, this study also evaluates the effect of spacing ratio between jets in predicting the flow field and jet behavior in near and far fields. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Heat transfer and pressure drop of supercritical carbon dioxide flowing in several printed circuit heat exchanger channel patterns

Closed-loop Brayton cycles using supercritical carbon dioxide (SCO{sub 2}) show potential for use in high-temperature power generation applications including High Temperature Gas Reactors (HTGR) and Sodium-Cooled Fast Reactors (SFR). Compared to Rankine cycles SCO{sub 2} Brayton cycles offer similar or improved efficiency and the potential for decreased capital costs due to a reduction in equipment size and complexity. Compact printed-circuit heat exchangers (PCHE) are being considered as part of several SCO{sub 2} Brayton designs to further reduce equipment size with increased energy density. Several designs plan to use a gas cooler operating near the pseudo-critical point of carbon dioxide to benefit from large variations in thermophysical properties, but further work is needed to validate correlations for heat transfer and pressure-drop characteristics of SCO{sub 2} flows in candidate PCHE channel designs for a variety of operating conditions. This paper presents work on experimental measurements of the heat transfer and pressure drop behavior of miniature channels using carbon dioxide at supercritical pressure. Results from several plate geometries tested in horizontal cooling-mode flow are presented, including a straight semi-circular channel, zigzag channel with a bend angle of 80 degrees, and a channel with a staggered array of extruded airfoil pillars modeled after a NACA 0020 airfoil with an 8.1 mm chord length facing into the flow. Heat transfer coefficients and bulk temperatures are calculated from measured local wall temperatures and local heat fluxes. The experimental results are compared to several methods for estimating the friction factor and Nusselt number of cooling-mode flows at supercritical pressures in millimeter-scale channels. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Experience in Remote Demolition of the Activated Biological Shielding of the Multi Purpose Research Reactor (MZFR) on the German Karlsruhe Site - 12208

In 2009, WAK Decommissioning and Waste Management GmbH (WAK) became owner and operator of the waste treatment facilities of Karlsruhe Institute of Technology (KIT) as well as of the prototype reactors, the Compact Sodium-Cooled Fast Reactor (KNK) and Multi-Purpose Reactor (MZFR), both being in an advanced stage of dismantling. Together with the dismantling and decontamination activities of the former WAK reprocessing facility since 1990, the envisaged demolishing of the R and D reactor FR2 and a hot cell facility, all governmentally funded nuclear decommissioning projects on the Karlsruhe site are concentrated under the WAK management. The small space typical of prototype research reactors represented a challenge also during the last phase of activated dismantling, dismantling of the activated biological shield of the MZFR. Successful demolition of the biological shield required detailed planning and extensive testing in the years before. In view of the limited space and the ambient dose rate that was too high for manual work, it was required to find a tool carrier system to take up and control various demolition and dismantling tools in a remote manner. The strategy formulated in the concept of dismantling the biological shield by means of a modified electro-hydraulic demolition excavator in an adaptable working scaffolding turned out to be feasible. The following boundary conditions were essential: - Remote exchange of the dismantling and removal tools in smallest space. - Positioning of various supply facilities on the working platform. - Avoiding of interfering edges. - Optimization of mass flow (removal of the dismantled mass from the working area). - Maintenance in the surroundings of the dismantling area (in the controlled area). - Testing and qualification of the facilities and training of the staff. Both the dismantling technique chosen and the proceeding selected proved to be successful. Using various designs of universal cutters developed on the basis of wall saws, both the activated steel liner and the inner reinforcing layer were cut remotely in one process. This allowed for the efficient execution of the following remote concrete removal steps using mining techniques. The electro-hydraulic demolition excavator that was purchased and then modified turned out to be an ideal tool carrier system with rapid-exchange coupling. Due to the high availability, no major delays occurred. This also was a result of the consistently implemented maintenance and repair concept. With the excavator installed in a modifiable scaffolding suspended from a rotating carrier ring, all dismantling areas could be reached and treated in spite of the small space. Thanks to an optimum organization of work-flows, routine change of dismantling work, and maintenance or repair, the iterative radiological measurement campaigns could be integrated in the whole activity without the dismantling work being disturbed significantly. The ventilation system with pressure grading and pre-filtration units ensured a low contamination level in the dismantling area. It was also possible to manage the dust formed by the milling of concrete surfaces. As it was possible to further cut metal parts and crushed concrete later on, residue flows were optimized. The planned overall period for testing, dismantling the bio-shield and removing the equipment was 36 months. The final duration was 39 months. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗