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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory importance of key DOE reactor research initiatives should be assessed early in the technology development process. Quality assurance requirements supportive of later licensing activities must also be attached to important research activities to ensure resulting data is usable in that context. Early regulatory analysis and licensing approach planning thus provides a significant benefit to the formulation of research plans and also enables the planning and development of a compatible AdvSMR licensing framework, should significant modification be required.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Advanced Demonstration and Test Reactor Options Study

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy”. Advanced reactors are defined in this study as reactors that use coolants other than water. Advanced reactor technologies have the potential to expand the energy applications, enhance the competitiveness, and improve the sustainability of nuclear energy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave similarly and retain the same ex-core features and characteristics. While INL has developed Designs A and B, LANL is evaluating a Hot Isostatic Pressed reactor configuration to overcome the drawbacks of the monolithic core design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO 2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO 2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave similarly and retain the same ex-core features and characteristics. While INL has developed Designs A and B, LANL is evaluating a Hot Isostatic Pressed reactor configuration to overcome the drawbacks of the monolithic core design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Special Purpose Application Reactors: Systems Integration Decision Support

Microreactors, also known as very-Small Modular Reactors (vSMRs) or Special Purpose Reactors (SPRs), are being considered for use in unique applications where other methods of megawatt level energy production are uneconomical or unavailable. For the purposes of this report, a microreactor is defined as meeting the following criteria: • Factory manufacturable • Easily transportable by truck, plane, train, and/or ship • Produce no more than 20 megawatt thermal (MWth) energy, in order to qualify as Hazard Category 2 under 10CFR830 • Maintains neutronic simplicity, allowing semi- or fully-autonomous operation In the United States (U.S.), there are two general types of microreactors under development: high-temperature gas reactors (HTGRs) and heat pipe reactors. These two reactor types may, due to their differences in working fluid temperatures, be more effectively suited to different use cases. Additionally, they will require different interfacing components depending on their intended use. This report identifies a broad range of use cases for micro-reactors. Additionally, multiple reactor technologies are examined with an emphasis on the reactor sub-systems and interconnections. Using this information, a decision framework is developed to guide future efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Reactor Fundamentals Handbook

The objective of the Knowledge Transfer Workshop is to transfer selected fundamental nuclear energy technology knowledge and experience to the nuclear industry’s next generation of leaders. The course provides the original bases for fundamental safety concepts, design features, and reactor technologies. The course discusses the “why” and the “how” for decisions that led to the current commercial nuclear reactor designs and regulatory framework. Various accidents and their attributes as well as other events that have strongly affected the evolution of the safety case, plant design, and operations are discussed. To prepare the course participant for the topics discussed, this handbook provides a crash course into the current state of commercialized Light Water Reactor technology. The material is introductory for the interested layperson and also serves as a refresher for scientists and engineers that may have studied Light Water Reactors at one time, but are currently not an expert in this field of nuclear engineering. This handbook is a brief tutorial and refresher training material on basic Light Water Reactor (LWR) core physics and plant systems. It is reference material for attendees of the Knowledge Transfer Workshop. There are two types of LWR technologies, Pressurized Water Reactor (PWR) and Boiling Water Reactors (BWR). This handbook discusses fundamental design traits of each providing contrasts between performance and safety features. The intent is to provide the participants with an understanding of nuclear terminology associated with LWR technology, including fundamentals, systems, safety, and regulations. This handbook is not an original textbook. Many authors have contributed collegiate level introduction to nuclear engineering texts. Rather, this material is intended to be a graded summary of important aspects of LWR reactor core and plant systems. It draws heavily on other resources. The interested reader is encouraged to follow citations made in this handbook to more thorough or detailed public references.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Preliminary Process and Instrumentation Design of Advanced Reactor Integration with Refineries and Hydrogen Production Facilities

With the ongoing push to decarbonize energy use and especially greenhouse gas emissions across all sectors, there are incentives to investigate how nuclear reactors may be used to generate clean energy and be used in various energy economies beyond just the electrical grid. Two initial integrations, high temperature steam electrolysis (HTSE) and oil refineries, are investigated in this first DOE Integrated Energy Systems (IES) program detailed industrial integration design report. Increasingly detailed reports are anticipated both for the industries discussed in this report and for additional industries in the future of the program. This report is a robust starting point showing how integration thermodynamic analysis establishes the requirements on the reactor and methods by which those requirements can be evaluated for specific reactor designs. Two Advanced Reactor Demonstration Program awardees are selected as representative designs for their respective technologies: NuScale for light-water reactors (LWRs) and X-Energy for high temperature gas reactors (HTGRs). Other reactor technologies or specific reactor configurations would require specific analysis similar to what is done in this report, thus this report can be a reference point by which to extend this work to other nuclear plant designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Measurement of neutron spectra in the experimental reactor LR-0

The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important task is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example. Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological research or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2). It is planned to create fully equipped complex facility possible to perform various experiments on the intensive neutron beam. Epithermal neutron beam enables development across the full spectrum of materials research for example shielding concrete tests or electronic devices construction improvement. Due to recent reports on the construction of the accelerator for the Boron Neutron Capture Therapy (BNCT) it has the opportunity to become useful and successful method in the fight against brain and other types of cancers not treated with well known medical methods. In Europe there is no such epithermal neutron source which could be used throughout the year for training and research for scientist working on BNCT what makes the stand unique in Europe. Also our research group which specializes in mixed radiation dosimetry around nuclear and medical facilities would be able to carry out research on new detectors and methods of measurements for radiological protection and in-beam (therapeutic) dosimetry. Another group of scientists from National Centre for Nuclear Research, where MARIA research reactor is located, is involved in research of gamma detector systems. There is an idea to develop Prompt-gamma Single Photon Emission Computed Tomography (Pg- SPECT). This method could be used as imaging system for compounds emitting gamma rays after nuclear reaction with thermal neutrons e.g. for boron concentration in BNCT. Inside the room, where H2 channel is located, there is another horizontal channel - H1 which is also unused. Simultaneously with the construction of the H2 stand it will be possible to create special pneumatic horizontal mail inside the H1 channel for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. It might expand the scope of research at the planned neutron station. Secondly it is planned to equip both stands with moveable positioning system, video system and facilities to perform animal experiments (anaesthesia, vital signs control, imaging devices, positioning). These all above make constructed station unique in the world (uranium fission converter-based beam) and the only one of such intense neutron beam in the Europe. Moreover implementation of the station would allow the development of research on a number of issues for researchers from all over the Europe. One of very important advantages of the station is undisturbed exploitation of the reactor and other vertical and horizontal channels. MARIA reactor operates 6000 hours per year and that amount of time will be achievable for research on the neutron station. It have to be underlined that new neutron station will work parallel to all another ventures. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Observation of Downward Facing Pool Boiling on a Hemispherical Vessel under External Reactor Vessel Cooling

It is a great challenge to ensure the safety of nuclear power plant during a severe accident. One of the major concerns is to overheat the reactor core which potentially may create a nuclear crisis if the core cooling fails. The concept of In-vessel retention (IVR) by passive external reactor vessel cooling (ERVC) in a flooded cavity of pressurized water reactor (PWR) is a viable and effective approach to retain radioactive core melt within the reactor pressure vessel (RPV). As long as the heat flux at the lower head of the RPV is lower than the critical heat flux (CHF) limit at the corresponding point, nucleate boiling will be the prevailing regime. In that case, the temperature of outer surface can be maintained near the saturated temperature of water such that the reactor vessel is able to be sufficiently cooled to sustain its thermal and structural integrity. Owing to the practical significance of this IVR-ERVC concept, there were many experimental and theoretical studies to investigate the two-phase flow natural circulation phenomena including the effect of inclination angle on the CHF, based on which many different empirical or semiempirical correlations were developed for CHF by fitting the experimental results. For the engineering applications, reactor-scale facilities such as ULPU series, KAIST facility, CYBL and FIRM were built to experimentally verify and validate the feasibility of ERVC for different design of reactors. Besides, for the research on physical mechanism of downward facing pool boiling, laboratory-scale experiments such as DELTA, CASA and SBLB were performed to visually and theoretically simulate the process of IVR-ERVC. All of the aforementioned test facilities and experiments provided a simulation of the downward facing boiling process and experimental study on the CHF limits during the ERVC conditions of different types of reactors. There were also some visualized investigations on the two-phase flow behavior in both quenching and steady state experiments. In this study, a scale-down three-dimensional downward facing pool boiling facility (iSBLB) was built to investigate the two phase natural circulation flow characteristics during the ERVC process. The general morphologies of vapor mass at different locations were visually recorded using the high-speed camera. The observation results showed that vapor behavior along the hemispherical curved surface was periodical. The boiling phenomenon was quite different from those on all other surface orientations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Modeling and simulation functional needs for molten salt reactor licensing

The modeling and simulation functional needs to license and design molten salt reactor concepts were assessed and an evaluation methodology was developed using a licensing-focused approach considering the modeling and simulation components that contribute to a molten salt reactor safety analysis report. These functional needs were prioritized by identifying the relationships between functional needs and using a quantitative weighting method to identify the most critical components to build a licensing case. This novel weighting method provided a list of functional needs prioritized relative to key advanced reactor licensing outcomes. In addition, functional need inputs were ranked to reveal those that feed the most modeling and simulation components. The findings from this prioritization effort were used to inform the modeling and simulation goals of the US Department of Energy Office of Nuclear Energy (DOE-NE) Molten Salt Reactor campaign.Modeling and simulation needs were grouped into three main categories: (1) source term definition, material accountancy, and radiological consequences, (2) safety analysis, and (3) steady-state neutronics and thermal hydraulics. Here, five near-term goals were identified among these focus areas that add value: (1) deploying a systems and safeguards analysis computational tool, (2) defining source terms for molten salt reactor analysis and interfacing with the US Nuclear Regulatory Commission on accident progression analysis, (3) developing point designs for liquid-fueled thermal-spectrum and fast-spectrum molten salt reactors, (4) building safety analysis specifications and reference solutions for Molten Salt Reactor Experiment transients, and (5) interfacing with the thermophysical and thermodynamic salt chemistry databases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Dismantling of primary shielding with prototype machines in a compact sodium cooled reactor - 15463

The Compact Sodium Cooled Reactor Facility Karlsruhe (KNK), a prototype reactor to demonstrate the Fast Breeder Reactor Technology in Germany, was in operation from 1971 to 1991. The dismantling activities started in 1991. The project aim is the green field in 2020. Most of the reactor internals as well as the primary and secondary cooling loops have already been dismantled. The contaminated sodium inventory has already been disposed of completely. Only the highly activated reactor vessel shielding structures remain. Due to their high dose rates, these structures must be dismantled remotely. For the removal of the primary shielding of the reactor vessel, 12 stacked cast iron blocks with a total mass of 90 Mg and individual masses of up to 15.5 Mg, a remote-controlled multifunctional dismantling device (HWZ) was designed, manufactured and tested in a mock-up. Since it is not possible to remove the segments in one piece for offsite treatment, a special prototype band-saw (BWZ) had to be developed for sawing the segments both horizontally and vertically into smaller pieces. The resulting sub-segments are transferred directly into repository containers. After the successful approval of the test sequences by the authorities and intensive training of the operational staff, both prototype machines were installed inside the reactor building containment for final assembly. The dismantling started in 2013. The present paper covers the successful dismantling of the primary shielding as well as the experiences gained through the development, use and improvement of these prototype machines. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

SCALE depletion capabilities for molten salt reactors and other liquid-fueled systems

Nuclear reactor systems that use fuel dissolved in a liquid have the potential for enhanced safety characteristics, improved fuel-cycle outcomes, and more efficient isotope-production configurations. In these reactor systems, the fueled liquid may simultaneously undergo irradiation, physical and chemical removal processes, and fueling. The modeling and simulation of this transmutation and decay with material additions and removals is an ongoing research area. An accurate simulation tool is critical to the reactor and fuel-cycle design, reactor deployment, and source-term characterization for these advanced reactor systems. The work described herein involved implementing, testing, and applying the capability to perform reactor physics simulations within the Oak Ridge National Laboratory-developed SCALE suite for nuclear systems analyses and design, leveraging much of its pedigree in quality-assurance and reactor-analysis capabilities. The functionalities to simulate irradiation with material feeds and removals had been added in ORIGEN, and the TRITON reactor physics sequence was extended to calculate the total removed material and track external nonirradiated mixtures to estimate separate processing or waste streams. Results from these capabilities align with analytical expectations obtained from ORIGEN for simplified test cases and with expectations for a molten salt reactor application. This implementation, available with the SCALE 6.3 release, provides for a more efficient and accurate material accountability methodology, allowing for the characterization, design, and analysis of the complete isotopic material inventory of advanced liquid-fueled systems for a variety of applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Sensitivity analysis, surrogate modeling, and optimization of pebble-bed reactors considering normal and accident conditions

This research provides a valuable tool that streamlines the optimization process while significantly increasing its accuracy. This study creates a robust framework for reactor design optimization by incorporating comprehensive modeling using the Comprehensive Reactor Analysis Bundle, or BlueCRAB, within the Multiphysics Object-Oriented Simulation Environment (MOOSE). BlueCRAB is the United States Nuclear Regulatory Commission's code suite for non-light water reactor analysis and includes the Griffin, Pronghorn, and Bison applications. This not only improves the efficiency of the optimization process but also enhances the reliability of the results. Such a tool is essential for advancing the state-of-the-art in pebble-bed reactor technology and is critical for achieving the goals of Generation IV reactors, which aim for safe, sustainable, and economically viable nuclear energy solutions. This work presents and applies this workflow on pebble-bed reactors while considering both normal and off-normal conditions. A representative gas-cooled pebble-bed reactor at equilibrium core conditions serves as the nominal design specification for normal operation and is based on previous research. The depressurized loss-of-forced-cooling accident is deployed for off-normal conditions in this work. After defining design-related parameters and quantities of interest regarding reactor safety and performance, this multiphysics model is sampled using the MOOSE stochastic tools module. The result is a comprehensive dataset of configurations, enabling sensitivity analysis and the generation of surrogate models. Subsequently, the dataset and surrogate models are employed in two optimization studies aimed at maximizing fuel utilization and economic profit while adhering to safety and operational constraints. Performing the optimization process with fuel utilization as the metric leads to an improvement of approximately 10%, compared to engineering-judgment-based nominal conditions. The optimization on economic profit leads to an estimated increase of ~300 million USD over the lifetime of the reactor.

97 MATHEMATICS AND COMPUTING↗