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The Super System Code of the Korea Atomic Energy Research Institute (SSC-K) has been developed for the transient analysis of the Korea Advanced Liquid MEtal Reactor (KALIMER) system. Recently, a detailed three-dimensional (3-D) core thermal-hydraulic model was developed to describe nonuniformities of radial temperature and flow within a subassembly and to decrease the uncertainties in the reactor safety margins during accident situations. The Shutdown Heat Removal Test-17 (SHRT-17) performed in the Experimental Breeder Reactor-II (EBR-II) and the postulated unscrammed events for the KALIMER conceptual design have been analyzed using a code system that has coupled a detailed 3-D core thermal-hydraulic model with SSC-K. The coupled code predicted behaviors for the experimental trends for the protected loss-of-flow SHRT-17. The KALIMER-150 design was adopted for a plant application of the same code system. Three events, unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), and unprotected loss of heat sink (ULOHS) were analyzed, and the simulation results were compared to those obtained using another code system that has coupled the Safety Analysis Section SYStem (SASSYS)-1 code with the same detailed 3-D core thermal-hydraulic model. The results, calculated with SSC-K coupled with the detailed 3-D core thermal-hydraulic model showed good agreement with the calculated results of the SASSYS-1 coupled code system for the UTOP and ULOF; however, some discrepancies were shown in the results for the ULOHS. These were found to have occurred because of a difference of the modeling for the decay heat removal system and primary coolant inventory. Through these analyses, the coupled code system was validated in order to be available for the safety analysis of a liquid-metal reactor (LMR) plant.

A heat transfer due to conduction through a coolant itself is not negligible in a liquid–metal cooled reactor (LMR). This portion of a heat transfer is frequently described with a conduction shape factor during the thermal-hydraulic design of an LMR. The conduction shape factor, which is highly dependent on a pitch-to-diameter ( P / D ) ratio, is defined as the ratio of the local conduction heat flux at a gap between two subchannels to the reference heat flux calculated by the averaged subchannel temperatures. The shape factors in heated triangular rod arrays for three different pitch-to-diameter ratios are generated through CFX calculations in the present study. The flow paths of 1.0–2.0 m in length are meshed into 180,000–360,000 volumes depending on the flow velocities. The SSG Reynolds stress model is used as a turbulent model in the calculations. The evaluated data fell between the heated-rod data and the plane-source data obtained by theoretical investigations. The conduction shape factors were found to be independent of the heating pattern of the rod arrays. Based on the evaluated data, a correlation for a liquid sodium coolant is suggested, which will improve the accuracy of the subchannel analysis codes for the thermal-hydraulic design of an LMR. When it is compared with the existing correlations, the suggested correlation is expected to enhance the reliability of the conduction shape factor because the data is evaluated by a more realistic numerical experiment.

The ability of point kinetics to describe dynamic processes in accelerator-driven systems (ADSs) is investigated. Full three-dimensional energy-space-time-dependent calculations, coupled with therm ...

One of the milestones in the roadmap of accelerator-driven transmutation of waste (ATW) of the U.S. Department of Energy is the design and construction of an accelerator-driven test facility (ADTF) with a thermal power of 100 MW. Analysis of the dynamic behavior of the ADTF has been carried out in the frame of a bilateral collaboration between the Forschungszentrum Karlsruhe and the Argonne National Laboratory (ANL). In the present study five different system configurations with various types of fuel and different types of coolant have been taken into consideration. In the systems with sodium as coolant, the transient behavior under the unprotected loss-of-flow scenario shows the most serious safety concern. As long as the external source is switched on, loss-of-flow will lead to an overheating of coolant, cladding and fuel. Boiling of coolant, cladding failure and molten fuel injection take place just in several seconds after the coast-down of the pump. Safety measures have to be designed for switching off the proton beam. In the system with liquid lead–bismuth eutectic (LBE) as coolant, the buoyancy effect is much stronger. Due to its high boiling point, coolant boiling and, subsequently, flow oscillation in fuel assemblies can be avoided. By a proper design of the heat removal system, the buoyancy-driven convection would provide a sufficiently high cooling capability of the reactor core, to keep the integrity of the fuel pins.

The applicability for inherent shutdown mechanisms in accelerator-driven systems (ADS) has been investigated. We study the role of reactivity feedbacks. The benefits, in terms of dynamics performan ...

The metalfuel version of the FPIN2 fuel element mechanics model has been incorporated into the SASSYS/SAS4A code system. In this implementation, SASSYS/SAS4A provides the fuel and cladding temperat...

This paper reports on a cooperative effort among European and U.S. analysts, which is an assessment of the comparative safety performance of metal and oxide fuels during accidents in a 3500-MW (thermal), pool-type, liquid-metal-cooled reactor (LMR) is performed. The study focuses on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower, and the unprotected loss-of-heat-sink (ULOHS). Core designs with a similar power output that have been previously analyzed in Europe under ULOF accident conditions are also included in this comparison. Emphasis is placed on identification of design features that provide passive, self-limiting responses to postulated accident conditions and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than do oxide-fueled reactors of the same design.

This paper summarizes an assessment of the regulatory framework and requirements for licensing a liquid metal reactor (LMR) for use in transmuting actinides, which was performed for the U.S. Department of Energy (DOE) Advanced Fuel Cycle Initiative (AFCI). Since the LMR designs currently under consideration are sodium-cooled, the assessment identifies and discusses requirements, issues, and topics important to the licensing process in general and those specific to sodium-cooled LMRs, as well as licensing options and associated recommendations. The goal of the regulatory framework assessment was to clarify and evaluate requirements that support the development of safe and cost-effective LMR designs. The scope of the assessment included an analysis of past and present licensing practices as well as an examination of possible future regulatory activities needed to support licensing LMR designs. Because this assessment included the identification of potentially problematic areas, a review of the past LMR licensing efforts was performed. Both technical and regulatory issues were identified and recommendations were made to address important issues. A review of the current regulatory framework for licensing a commercial reactor and the associated licensing schedules was performed as part of the assessment. In addition, specific options proposed by the U.S. Nuclear Regulatory Commission (NRC) for licensing an LMR were also assessed with regard to their potential impacts on different stakeholders, which include the NRC, DOE, industry, and the public. In addition to the licensing of a commercial LMR, the assessment also identifies and evaluates licensing options for an LMR prototype. The regulatory assessment supports a conclusion that a safe, licensable LMR design is fully feasible. The knowledge applied in the LMR design will be reinforced by past experience and available technology. The licensing of an LMR is expected to be manageable, notwithstanding the uncertainties associated with regulatory, technical, and other issues. With forward-looking planning, effective management, and adequate resources, the process of obtaining a license for an LMR would be greatly facilitated.

This paper characterizes the areas of Integral Fast Reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure of critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR Safety and related Base Technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorites.

The liquid-salt-cooled very high-temperature reactor (LS-VHTR), also called the Advanced High-Temperature Reactor (AHTR), is a new reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. Depending upon goals, the peak coolant operating temperatures are between 700 and 1000°C, with reactor outputs between 2400 and 4000 MW(t). Several fluoride salt coolants that are being evaluated have melting points between 350 and 500°C, values that imply minimum refueling temperatures between 400 and 550°C. At operating conditions, the liquid salts are transparent and have physical properties similar to those of water. A series of refueling studies have been initiated to (1) confirm the viability of refueling, (2) define methods for safe rapid refueling, and (3) aid the selection of the preferred AHTR design. Three reactor cores with different fuel element designs (prismatic, pebble bed, and pin-type fuel assembly) are being evaluated. Each is a liquid-salt-cooled variant of a graphite-moderated high-temperature reactor. The refueling studies examined applicable refueling experience from high-temperature reactors (similar fuel element designs) and sodium-cooled fast reactors (similar plant design with liquid coolant, high temperatures, and low pressures). The findings indicate that refueling is viable, and several approaches have been identified. The study results are described in this paper.Copyright © 2006 by ASME

An assessment has been performed for an Accelerator Transmutation of Waste (ATW) concept based on the use of the high temperature gas reactor technology. The concept has been proposed by General Atomics for the ATW system. The assessment was jointly conducted at Argonne National Laboratory (ANL) and Los Alamos national laboratory to assess and to define the potential candidates for the ATW system. This report represents the assessment work performed at ANL. The concept uses recycled light water reactor (LWR)-discharge-transuranic extracted from irradiated oxide fuel in a critical and sub-critical accelerator driven gas-cooled transmuter. In this concept, the transmuter operates at 600 MWt first in the critical mode for three cycles and then operates in a subcritical accelerator-driven mode for a single cycle. The transmuter contains both thermal and fast spectrum transmutation zones. The thermal zone is fueled with the TRU oxide material in the form of coated particles, which are mixed with graphite powder, packed into cylindrical compacts, and loaded in hexagonal graphite blocks with cylindrical channels; the fast zone is fueled with TRU-oxide material in the form of coated particles without the graphite powder and the graphite blocks that has been burned in the thermal region for three critical cycles and one additional accelerator-driven cycle. The fuel loaded into the fast zone is irradiated for four additional cycles. This fuel management scheme is intended to achieve a high Pu isotopes consumption in the thermal spectrum zone, and to consume the minor actinides in the fast-spectrum zone. Monte Carlo and deterministic codes have been used to assess the system performance and to determine the feasibility of achieving high TRU consumption levels. The studies revealed the potential for high consumption of Pu-239 (97%), total Pu (71%) and total TRU (64%) in the system. The analyses confirmed the need for burnable absorber for both suppressing the initial excess reactivity and ensuring a negative temperature coefficient under all operating conditions. Additionally, current results suggest that it may be preferable to use a double strata thermal critical system and fast subcritical system to achieve nearly complete destruction of the TRU oxide fuel. The report gives a general description of the system proposed by General Atomics. The major design parameters (degrees of freedom), which can be altered to optimize the system design, and the constraints, which guide the design and the optimization studies are described. The deterministic and the Monte Carlo neutronics codes and models used for the neutronics analysis and assessment are presented. The results of fuel block and whole-core parametric studies performed to understand the physics are given including the effect of various fuel management schemes on the system performance. A point design is described including the system performance results for a single-batch and three-batch loading schemes. The major design issues, which need to be addressed during further studies, are discussed.

Transient safety characteristics of accelerator-driven systems using advanced minor actinide fuels have been investigated. Results for a molybdenum-based Ceramic-Metal (CerMet) fuel, a magnesia-bas ...