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Initial Assessment of TeF 6 Adsorption

Research and development that supports the management of off-gases from nuclear fuel reprocessing has historically been focused on the off-gas streams that arise from aqueous reprocessing technology. However, as Gen-IV reactor development pathways move toward deployment, alternative spent nuclear fuel (SNF) processing and disposition pathways have been considered more actively. This work is focused upon aspects of fluoride volatility (FV) processing. The versatility of this method for deployment against multiple types of spent fuel encourages the continued advancement of both the primary separations processes and the secondary processes, including waste treatment, material control and accountability, and engineering designs. Recent work noted that TeF 6 , the most highly volatile fluorinated compound produced during FV processing, did not have a clear abatement technology recommended in the literature. Thus, this work performed scoping tests on activated alumina and copper shot to assess whether they could be used to remove TeF 6 from gas streams that bear F 2 . Previous work on this topic was not well described in the literature and was not performed with excess F 2 in the stream as would be typical of spent fuel processing via FV. To support an understanding of the concentration of TeF 6 contacting the adsorbent beds, a series of preliminary testing identified the TeF 6 production rate and the equilibrium concentration of TeF 6 in the gas stream contacting the adsorbent. Excess F 2 was determined to not affect the ability of activated alumina to remove TeF 6 from the gas stream quickly and completely. A determination of whether excess F 2 affected the distribution depth of Te in the sorbent bed is still pending analysis of the used sorbent. Literature suggests that when activated alumina is near saturation, TeF 6 could migrate from the sorbent bed. Future testing should investigate this possibility. Copper metal did not adsorb TeF 6 in the presence of F 2 across the sorbent temperature range of 50 to 335°C. F 2 was fully removed by the copper bed. Although preliminary thermodynamics would indicate adsorption to be energetically favorable, other factors that impact adsorption (e.g., slow kinetics, excess fluorine on the copper surface, an unfavorable transition state) are likely preventing the adsorption of TeF 6 at an easily measurable rate. The work also investigated TeF 6 production from multiple forms of Te in the temperature range of 100 to 250°C. No previous study had assessed the initial reaction rates for this process. A carefully designed study allowed determination of the activation energy for the production of TeF 6 from Te metal. The substantial amount of data collected during this study merits analysis beyond what is described here. A more in-depth kinetic analysis will be pursued. Additional analytical results will provide the ability to benchmark adsorption coefficients for TeF 6 , understand the distribution of TeF 6 within the alumina bed, and better understand the effect of F 2 partial pressure on Te fluorination. The data from this report, as supplemented by these additional analyses, will be submitted to a peer-reviewed journal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Enabling Load Following Capability in the Transatomic Power MSR

This project is dedicated towards designing a fuel processing system that enables liquid-fueled molten salt reactors (MSR) to load follow by removing the dissolved xenon in the fuel salt. As one of the Gen-IV nuclear reactor concepts, the molten salt reactor receives increasing development interests in the recent years. One distinguishing feature of the liquid-fueled molten salt reactor is its improved ability to operate in a load-following mode by including the unique online fission product removal system. Load-following means that the reactor changes its power output based on the demand on the grid. Most of the current operating nuclear reactors have limited load-following ability and operate as the base load on the grid. Due to the rapid increase of solar energy, the requirement on load-following capacity is significantly increased because of the varying power output of the solar panels, yet the traditional load-following capacity is expected to decrease as the decarbonization of the grid continues. Therefore, the ability to perform load-following operation for the nuclear reactors will greatly enhance the resilience of the grid and make nuclear energy more economically competitive. This load-following feature is included in many commercial molten salt reactor designs, such as the designs by Transatomic Power, Terrestrial Energy, and Flibe Energy. Unfortunately, detailed analysis of the fuel processing system for commercial scale MSRs is still lacking, as well as how the fuel processing quantitively impacts the load-following operation. Moreover, experimental data for many of the underlying physics of fuel processing is limited. This project aims to pave the way for the fuel processing technology to advance to the commercial stage by performing combined experimental and simulation research. During the project period, four interconnected aspects of the development of the fuel processing system in liquid-fueled molten salt reactors are investigated. These aspects are the simulation and analysis of the fission product removal system, the fuel cycle simulation, the coupled reactor neutronics and thermal hydraulics transient simulation, and the gaseous fission product removal experiment. Multiphase CFD simulations are performed for components of the processing systems, and simplified air-water experiments are carried out to provide validation data. It is concluded that the CFD simulation can satisfactorily predict the system level performance of the components, and engineering models are constructed based on this success. Fuel cycle analysis is performed for two representative MSR design, the MSBR and the Transatomic Power MSR. Open-source code SaltProc is developed to incorporate the unique fuel processing system of the MSRs. It is concluded that the removal of xenon is essential for load-following operation in thermal spectrum MSR and Molten Salt Breeder Reactor. For the Transatomic Power MSR, the xenon poisoning effect is negligible due to its relatively fast neutron spectrum, though the overall fuel cycle economics still benefits from the removal of xenon. Coupled reactor neutronics and thermal hydraulics transient simulation is performed specifically for the Transatomic Power MSR. It is concluded that the reactor core design could perform power ramping fast enough to satisfy load-following operation. Combining the findings from each aspect, it is concluded that the load-following operation of a thermal neutron MSR is dependent upon the removal of xenon, which could be achieved for a commercial sized reactor using continuous inert gas sparging in a separate system with reasonable dimensions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Strategic Plan: Decrease Critical Minerals Waste Through Enabling Advanced Manufacturing Techniques

In this report we developed a strategic outlook derived from previous findings by our Phase-2 Scorecard Report and the Critical Mineral Evaluation Report. In addition, Phase 2 Materials Scorecards have been developed in 2022 to provide a detailed justification and traceability for the adoption of additively manufactured structural materials and ceramics for nuclear deployment. The report provides a strategic vision to harvest the benefits of reduced material and energy demands by enhanced deployment of advanced manufacturing in nuclear technology. The objective of this report is to address the needs for imminent experimental/computational studies to optimize the deployment of advanced manufacturing for its energy-, cost-, and critical materials saving potentials. We display what further actions are needed to minimize supply risks of critical minerals and to lower the economic vulnerability when implementing the additional nuclear capacity projected by 2050, in support of the U.S. goal of net-zero carbon emission. The critical minerals essential for Gen-IV nuclear technology show either low disruption potential (supply risks) combined with high economic vulnerability, such as: copper, nickel, and chromium; or low supply risks combined with medium economic vulnerability, such as: manganese, molybdenum, tungsten, vanadium, tantalum, and zirconium. Enhancements in quality and yield of critical mineral recycling and the recovery of critical minerals from solid and industrial waste streams are needed to enhance supply security for stainless-steel production and to shift the time to scarcity for society to future years.

36 MATERIALS SCIENCE↗

Coupling of Pronghorn and RELAP-7 for a Pebble Bed Reactor

High temperature gas cooled reactors (HTGR) are a candidate for timely Gen-IV reactor technology deployment because of high technology readiness and walk-away safety. Among HTGRs, pebble bed reactors (PBRs) have attractive features such as low excess reactivity and online refueling. Pebble bed reactors pose unique challenges to analysts and reactor designers such as continuous burnup distribution depending on pebble motion and recirculation, radiative heat transfer across a variety of gas-filled gaps, and long design basis transients such as pressurized and depressurized loss of forced circulation. Modeling and simulation is essential for both the PBR’s safety case and design process. In order to verify and validate the new generation codes the Nuclear Energy Agency (NEA) Data bank provide a set of benchmarks data together with solutions calculated by the participants using the state of the art codes of that time. An important milestone to test the new PBR simulation codes is the OECD NEA PBMR-400 benchmark which includes thermal hydraulic and neutron kinetic standalone exercises as well as coupled exercises and transients scenarios. In this work, the reactor multiphysics code MAMMOTH and the thermal hydraulics code Pronghorn, both developed by the Idaho National Laboratory (INL) within the multiphysics object-oriented simulation environment (MOOSE), have been used to solve Phase 1 exercises 1 and 2 of the PBMR-400 benchmark. The steady state results are in agreement with the other participants’ solutions demonstrating the adequacy of MAMMOTH and Pronghorn for simulating PBRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

HTGR Simulation Methods & International Collaborations

ART-GCR “Methods” activity is split between Experimental Validation data from the ANL NSTF and OSU HTTF (next three presentations). HTGR core simulation (this presentation). International collaboration within OECD Generation-IV (Gen-IV) and USA/Japan bi-lateral agreements (this presentation) HTGR Simulation Methods No new NE-52 funding for HTGR Methods support in FY20; ~$200K FY19 carry-over funds only. Consists of international code-to-code benchmarks (IAEA CRP on HTGR UAM and OECD/NEA MHTGR-350) and refinement of a few-group Pebble Bed Reactor (PBR) cross section (XS) generation methodology. Funding will be requested in FY21 to produce the final reports for the two benchmarks and continue the development of the PBR XS generation methodology. Additional (non-ART) HTGR-related support work at INL NEAMS: HTR-Application work package at INL Create a benchmark for the pebble shuffling and depletion algorithms being developed for NEAMS Griffin code. iFOA award with X-Energy: Develop independent Monte Carlo model of Xe-100 design. Independent design confirmatory analysis of Xe-100 design using NEAMS tools Griffin and Pronghorn. Support X-Energy design team to use their own legacy design tools (VSOP99 and MGT). Support NEAMS Griffin and Pronghorn development team for the iFOA needs (received $50K additional funding for required development).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Modeling The DLOFC Accident Scenario of HTR-PM Equilibrium Core Using NEAMS Tools

High-Temperature Gas-cooled Reactors (HTGRs) have excellent characteristics in terms of safety and high thermal efficiency, and they are gaining a large interest from the industry as a candidate of Gen-IV reactors for a wide range of applications. The High Temperature gas-cooled Reactor Pebble-bed Module project (HTR-PM) is one of these designs and where helium gas is used to cool the pebble-bed region that consists of spherical fuel elements moderated with graphite. The HTR-PM design is based on the combined experience from the German pebble-bed reactor program from the 1960s through the 1990s and the HTR-10 experience in China during the 2000s. Idaho National Laboratory has a long experience in modeling of HTGRs working in developing neutronics and thermal hydraulics tools for the proper modeling of these reactors. The neutronics code Griffin has the capability to model pebble depletion . While the thermal hydraulics code Pronghorn was developed mainly to model the pebble bed reactors with the porous media assumption. In this work, an equilibrium core Multiphysics model was developed for the HTR-PM reactor to analyze the depressurized loss of forced cooling accident scenario (DLOFC). This paper is organized as follows: First, a brief description of the reactor and model specifications are provided. Then, the developed Multiphysics model is discussed. Finally, verification results of the steady-state equilibrium core and DLOFC accident are presented followed by a summary of the conclusions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Overview of PGSFR Project

The long-term plan for advanced SFR development was authorized by Korean Atomic Energy Commission (KAEC) in 2008 and updated in 2011, which will be carried out toward the construction of a SFR prototype plant called PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) by 2028. The primary purpose of PGSFR is to demonstrate the technology reducing radioactive waste of spent nuclear fuel from nuclear power plants (NPP) in Korea. There are 24 NPPs in operation nationwide covering 30% of domestic electricity demand. The spent nuclear fuel is currently stored in the on-site storage pool of NPPs. Then, it has become a critical issue to ensure technical options to handle spent nuclear fuel in the long term basis to secure the sustainable operation of NPPs. PGSFR is a pool type sodium-cooled fast reactor with the power of 150 MWe or 400 MWth. PGSFR uses metallic fuel to make use of inherent safety characteristics of metallic fuel. The spent nuclear fuel will be processed using pyro-process to produce U-TRU-Zr fuel for PGSFR. But the initial core shall be loaded by low enriched U-Zr fuels. PGSFR has two intermediate loops with two intermediate heat exchangers (IHX) for each loop to deliver the generated heat from the core to power conversion system. There a steam generators producing super-heated steam in each loop. There are four decay heat removal circuits; two passive and two active ones. The reactor building is built on a seismic isolation base mat which can significantly mitigate seismic vibration of major components in the reactor building. The reactor building has a containment structure to isolate the reactor building under specific accident conditions and also provide structural integrity at the time of aircraft impact accident. PGSFR project is the national development program of Korea to construct the first SFR in Korea by 2028. The primary goal of PGSFR is the demonstration of reduction of radioactive waste from spent nuclear fuel by incineration in PGSFR. SFRA will flexibly cope with the licensing environment to minimize the uncertainty in licensing and safety concerns. The successful construction of PGSFR will bring Korean nuclear industry a step closer to guarantee the sustainable operation of NPPs nationwide and finally most importantly will solve the spent nuclear fuel issue of Korea. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

PGSFR Core Design and Performance Characteristics

Korea Atomic Energy Research Institute (KAERI) has developed a Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) as a representative of Gen-IV concepts. The final PGSFR core will be composed of transuranic (TRU) fuels to transmute LWR spent fuels and self-recycled its fuels, in combination with a pyro-processing facility. However, the initial POSFR core consists of low enriched uranium (LEU) fuels in order to: 1) provide an irradiation test for TRU fuels and structural materials, 2) keep pace with the development of pyro-processing technology, and 3) obtain operational experience on the reactor systems and components. During the first phase of PGSFR development plan (2012-2015), the power level was decided to be 150 MWe from many trade-off studies using different power level. The PGSFR uranium core performance characteristics at the initial, transition and equilibrium states have been carried out. In this paper, the PGSFR core design and related performance characteristics are introduced. The PGSFR uranium core design was performed at the initial/transition/equilibrium core based on the candidate fuel management scheme. It was confirmed that this design can satisfy surely all design limit and requirement. The chapter 4 of PGSFR Preliminary Safety Information Document (PSID) published in December 2015 contains the detailed information and characteristics of PGSFR core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Safety Analyses of Design Basis and Design Extended Condition Events for the PGSFR

KAERI (Korea Atomic Energy Research Institute) has been developing the PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor), which is a pool type sodium-cooled fast reactor with the thermal power of 392.2 MWt, uses metallic fuel of U-10%Zr for a core. The preliminary specific design of PGSFR is completed in 2015. The main parts of PGSFR consist of PHTS (Primary Heat Transport System), IHTS (Intermediate Heat Transport System), two SG (Steam Generator), and DHRS (Decay Heat Removal System). The main feature of PGSFR in safety is from the adoption of metallic fuel core, leading to inherent reactivity feedback mechanisms and high thermal conductivity. In this work, safety analysis on DBE (Design Basis Event) and DEC (Design Extended Condition event) for PGSFR preliminary specific design is conducted using MARS-LMR code. Conservative and best-estimated approaches are applied to DBA and DEC, respectively. It is found that the most severe accidents in the AOO, DBA class 1, DBA class 2, and DEC are LOF, OPS, TOP and ULOF event, respectively. And all analysis results are found to meet safety acceptance criteria requirements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

R and D activities on VHTR materials at KAERI - 18496

KAERI started VHTR materials R and D in 2006 as part of its role in supporting the development of key technology for the nuclear research and development program. Over the past ten years, research has been conducted on various materials including Alloy 617 for an IHX and hot gas duct, 9Cr-1Mo for an RPV, graphite for a reflector and the support structures in the core region, Cf/C and SiCf/SiC composites for a control rod, and Fe-Si, SiC, Hastelloy{sup R}, and other novel metals for an SI process reactor. Mechanical properties including creep and fatigue, environmental effects, and irradiation behaviors have also been heavily evaluated as part of the materials research. Most of the results have been contributed to the GEN-IV VHTR materials collaboration. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Helium Brayton Cycle Design for Advanced High Temperature Reactor System

Gen-IV reactors have many advantages over current generation reactors. Higher outlet temperatures in these next generation reactors allow for other power cycles to be considered, such as the helium Brayton cycle. Additionally, the higher temperatures allow the energy to be utilized in other industrial applications. The advanced high temperature reactor concept utilizes the helium Brayton cycle for electricity production. The system consists of four components shown in Fig. 1. This includes a turbine, recuperator, precooler, and compressor. The helium coolant exits the secondary heat exchanger (SHX) at a high temperature and enters the turbine. Helium spins the turbine causing the generator to spin producing electricity. The pressure and temperature drop as the coolant travels through the turbine. The helium enters a recuperator where the cold side stream exiting the compressor is heated up. Helium then travels through a precooler where water is used to cool down the helium before entering the compressor. Helium enters the compressor where the temperature and pressure of the coolant increase. Compressed helium travels through the recuperator heating up to the desired temperature before entering the SHX again. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

SAS4A model development for the analysis of postulated severe accidents in metal fuel sodium fast reactors

The SAS4A code has been designed to simulate the events that occur in Sodium Fast Reactor (SFR) during the initiating phase of severe unprotected accidents. The initial SAS4A models were focused on the behavior of oxide fuel cores. An effort to expand the SAS4A models for the analysis of metal fuel cores is currently underway. This paper describes the status of the SAS4A metal fuel model development and presents initial results from the simulation of a postulated fuel assembly blockage accident in the Prototype Gen-IV Sodium Fast Reactor (PGSFR). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Analysis on the Heat Transfer Phenomena Inside the Scaled-Down VHTR RCCS Riser Experiment

Very High Temperature gas-cooled Reactor (VHTR) that is one of GEN-IV reactors has been developed for the purpose of hydrogen production. VHTR is characterized by high operating temperature and stepped-up inherent safety given by passive safety systems such as Reactor Cavity Cooling System (RCCS). In case the active heat removal system losses its capability, RCCS significantly contributes to the inherent safety of VHTR by removing the decay heat transferred from the reactor core to the reactor vessel. Therefore, the performance of RCCS is the key parameter to ensure the safety of the reactor. Several computation and experiment researches on the performance of RCCS have been conducted. There are two types of RCCS depending on its coolant; one is the air-cooled type and the other is water-cooled type. This study focused on the former type. Due to the excessive height of the prototype RCCS, experiment facilities were constructed in reduced scale. Korea Atomic Energy Research Institute (KAERI) constructed the quarter-scale RCCS experiment facility, NACEF and Argonne National Laboratory (ANL) has constructed half-scale experiment facility, NSTF. University of Wisconsin (UW) also has the quarter-scale experiment facility and Texas A and M University has conducted separate effect test on the jet effect at the upper plenum of RCCS. From the reduced scale experiment result, the performance of RCCS such as a heat removal rate of RCCS and a peak temperature of reactor vessel were evaluated. When the reduced-scale result is applied to the prototype, however, a distortion could be brought about due to the reduced height of RCCS in the experiments. Especially, if the heat transfer regime inside the riser which roles as a heat exchanger to discharge the decay heat into the atmosphere is not identical, the riser wall temperature and the consequent reactor vessel temperature of the prototype could be incorrectly predicted. Thus, the convective heat transfer phenomena inside the riser duct both in the reduced scale and in the prototype is necessary to be investigated. In this study, the preliminary analysis on the heat transfer phenomena inside the RCCS riser was conducted. The subjects of the analysis were the prototype RCCS riser of PMR 200, and its scaled-down experiment facilities, NACEF. First, the geometry, flow and thermal conditions of the riser in PMR200 RCCS and NACEF were introduced with a scaling method. And then, the heat transfer regimes and the heat transfer coefficient corresponding to each condition were predicted using correlation and CFD calculation. Finally, planned activities including the experiment was introduced. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Electrochemical Behavior of Lanthanum in Molten LiF-NaF-KF 'Flinak' Eutectic Salt

In order to develop the Fluoride High- Temperature Reactor, an emerging Gen-IV reactor concept, a primary coolant cleanup system must be developed to allow online reprocessing of the primary molten fluoride salt. This research is focused on the determination of electrochemical properties of rare earths (REs) in molten fluoride salt related to the FHR. The FHR utilizes two fluoride salts. These fluoride salts are LiF-NaF-KF (46.5-11.5-42 mol%), melting point 454 deg. C, also known as 'Flinak', and LiFBeF{sub 2} (33-67 mol%), melting point 459 deg. C, also known as 'Flibe'. Flibe is selected as the primary coolant salt due to its low neutron cross-section (Li{sup 7}-enriched), superior heat transfer properties, and low melting point. However, due to the beryllium content in Flibe, it is highly toxic to humans and Flinak is preferred for the secondary and tertiary coolant salts. The work presented here is specifically focused on the behavior of lanthanum trifluoride (LaF{sub 3}) in molten Flinak. Furthermore, updated data and analysis regarding the electrochemical behavior of dysprosium trifluoride (DyF{sub 3}) is also presented. Previous work related to this topic was presented at the 2016 ANS winter conference. The goal of this research is to develop a contaminant cleanup system for the primary loop of the FHR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Preliminary Development of a Nuclide Inventory Code for HTGR

Nuclide inventory calculation for a reactor core is the basis of radiation protection, shielding design and radioactive environmental impact assessment of the power plant. HTR-PM (high-temperature gas-cooled reactor pebble-bed module) under construction in China is one of the candidates for Gen-IV advanced nuclear power system.1 In the design stage of HTR-PM, KORIGEN has been utilized in the inventory calculation.2 However, the current version of KORIGEN was developed several decades ago. In order to improve the design capability of the analysis code package for HTR-PM, a new computer code for nuclide inventory calculation called NUIT (Nuclide Inventory Tool) is being developed based on the DEPTH3 code. The NUIT code tries to utilize accurate and efficient burnup solvers, incorporate well-validated and most recent nuclear data libraries, and provide user-friendly IO interface. This paper presents the preliminary work in the development of NUIT.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Study on Neutronics Calculation Method of SCWR with Annular Fuel Assembly

Supercritical Water-Cooled Reactor (SCWR) is one of Gen-IV reactor systems for its advantages over LWRs, such as high thermal efficiency, simplified reactor system and economics. Annular fuel was proposed by MIT to supplant conventional solid fuel and improve power density in LWRs. Annular fuel is cooled from both inside and outside of the fuel. Thus a lower fuel temperature and cladding surface temperature can be achieved with higher power density. SCWRs will benefit from this advantage because high cladding temperature (usually over 650 degree C) is one challenge of SCWRs. Using annular fuel in SCWRs is very promising. This is also because the inner hole of the annual fuel can provide extra neutron moderation. Supercritical water density is too low to provide enough moderation. Different analysis codes have been developed all over the world to carry out SCWR pre-conceptual core design. Yamaji et al. developed SRAC/ASMBURN/COREBN to study Japan version of SCWR. Monti et al. used ERANOS to analyze HPLWR (High Performance Light Water Reactor). WIMS-D4/CITATION was also used to calculate HPLWR by Tashakor et al.. These codes employ a two-step method used in PWRs calculation. Lattice/Assembly calculation is carried out as the first step. In the second step, few group constant of different fuel assembly types is then used in core calculation. MCNP had also been used to study SCWR by Watta et al.. Due to large computation time cost, it's used mostly to analyze fuel assembly or core performance rather than pre-conceptual design, in which much optimization is needed. In this research, a neutronics calculation code for SCWR is developed using two-step method. This code is verified by comparing with Monte-Carlo code. Fuel assembly homogenization method is studied to improve the code accuracy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Methodology of Experiment Design for Improving Transient Critical Heat Flux Model Development

One phenomenon that contributes to nuclear fuel and clad failure mechanisms is the departure from nucleate boiling (DNB). Accident scenarios such as reactivity-initiated accidents (RIAs) provide conditions where the frequency of boiling crisis is much more likely to occur and on much shorter time scales than loss of coolant (LOCA) or loss of flow (LOFA) accidents. Here, boiling crisis in the form of DNB is attributed to transient-induced heat flux values that exceed the critical heat flux (CHF). The resulting effects, such as bubble crowding and eventual dryout, can affect flow boiling heat transfer and compromise the structural integrity of nuclear reactor materials. The focus of the work proposed here lies in developing a rigorous process using out-of-pile testing, under simulated RIA conditions, to determine the transient critical heat flux of cladding materials, without contributions from aforementioned radiation-induced effects that have been observed during in-pile experiments. This study provides a road-map for developing and performing experiments that simulate the power profile and energy deposition aspects of transient events, specific to RIAs, on the laboratory scale. The goal of our project is to observe and measure DNB in an annular flow channel around a single heater rod that is operating and nominal thermal hydraulic conditions of a PWR. It is also an aim of these experiments to determine how to quantify the gap between SS and transient CHF cases, based on state-dependent properties of the cladding material type that is being interrogated. Traditionally, CHF is determined as a combination of four factors: hydraulic diameter, system pressure, mass flow, fluid quality. The authors propose that an additional measureable quantity, which is dependent on the structural properties of the cladding material itself, is partly reasonable for the difference between measured and model-developed CHF values. Using a carefully metered approach will allow for several key concepts to be realized: compensation for transient-based effects to current models for CHF; adaptation of current models for different cladding material types, including Gen-IV concept materials; and implementation of a correction factor to account for both power transient-based and radiation-induced effects on the CHF that have not yet been accounted for.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Need for research and training reactors for advanced reactor designs

Full text of publication follows. Research and training reactors have served a valuable role in helping train workforce for currently operating fleet of light water reactors. These research and training reactors have been used in reactor laboratory classes to familiarize the students with such vital concepts as approach to criticality, reactor period, neutron moderation, reactivity, flux distribution and leakage, etc. As the industry moves toward advanced non-light-water reactor designs, it is critical that research and training reactors be developed and deployed at university campuses to help train the new generation of nuclear and non-nuclear engineers who are likely to design, build, and operate these advanced reactors. Among the designs currently being pursued for nuclear power generation include molten salt, sodium cooled, and gas cooled designs, with options for various fuel forms. Thus, industry and DOE in collaboration with academic institutions should devise plans on how to familiarize the next generation of nuclear workforce with hands-on experience necessary for such designs. These research and training reactors will play a vital role in familiarizing the future workforce with hands-on experience with concepts associated with fast spectrum reactors, gas cooled reactors, and other features not associated with light water reactors. In addition to classical nuclear engineering concepts, these advanced research and training reactors can also be used for hands-on training as well as for research on features being considered in the design of GEN-IV reactors: cyber security for digital control room operations, hybrid energy system, hydrogen generation, district heating, autonomous control... (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗