Depressurized Loss Of Forced Cooling Research Articles

Investigating geometry adjustments for enhanced performance in a PeLUIt-10 MWt pebble bed HTGR with OTTO refueling scheme

This research investigates the impact of changes in reactor geometry to discover more optimal dimensional options for nuclear power plants with potential applications in remote areas of Indonesia. The reactor used is PeLUIt-10, a pebble bed High-Temperature Gas-cooled Reactor (HTGR) with a 10 MWt capacity and a “once-through-then-out” or OTTO fuel loading scheme. The neutronic and thermal–hydraulic performances of PeLUIt-10 were analyzed with its core size altered, first by modifying its height-to-diameter ratio (H/D ratio) and second by reducing its core volume. These analyses were performed using the Pebble Bed Reactor Neutron Diffusion (PEBBED) Code, a specialized tool to analyze HTGR reactor physics parameters, particularly Pebble Bed Reactors.. Neutronic parameters analyzed include total fuel flow, burnup, power peaking factor, and power density distribution. For thermal-hydraulics and safety, parameters include steady-state and transient fuel temperatures, especially in Depressurized Loss of Forced Cooling (DLOFC) accidents. Results show that the optimal design, maintaining a volume of 5 m3, is a reactor with a height of 159.57 cm and a diameter of 200 cm (H/D ratio of 0.8). The fuel can achieve a maximum burnup of 80.97 MWD/kg-HM at this size. Power density distribution at these dimension is better than another dimensions, with a relatively low power peaking value of 1.4. For safety parameters, the fuel temperature in transient conditions remains below the safety limit. Meanwhile, when the core volume is reduced, the minimum burnup target of 60 MWD/kg-HM at a reactor volume of 4.4 m3 with a diameter of 172 cm and height of 189.2 cm. Thus, for OTTO-cycle PeLUIt-10, altering H/D ratio is more beneficial.

Multiphysics Analysis of PLOFC and DLOFC Accidents in the HTGR-TR

High-temperature, gas-cooled reactors (HTGRs) are promising advanced reactor designs. Leveraging passive safety features, higher core outlet temperatures, previous commercial operational experience, and potential coupling with nonelectrical systems, HTGRs present many advantages over traditional light water reactors and other advanced reactor designs. The High-Temperature, Gas-Cooled Test Reactor (HTGR-TR) is a point design developed by Idaho National Laboratory as a response to the need for the expansion of U.S. test reactor capabilities. The HTGR-TR was designed to serve as a test reactor for Generation-IV small modular prismatic block HTGRs and to provide irradiation data, two crucial contributions to the development of advanced reactors. As in any reactor design, it is necessary to understand the behavior of the reactor during accident scenarios to determine the overall safety of the design. Pressurized loss of forced cooling (PLOFC) and depressurized loss of forced cooling (DLOFC) are two accidents that can occur in HTGRs and that present challenges to decay heat removal. In order to understand the behavior of the fuel during these accidents, the resulting mechanical behavior and fission product migration in the fuel are evaluated. Using the HTGR-TR design, which utilizes tristructural isotropic (TRISO) fuel, a multiphysics analysis during the PLOFC and DLOFC transients was performed. This study utilizes neutronics, thermal hydraulics, and fuel performance modeling to analyze the behavior of TRISO particles during steady-state operation and the described transients. This analysis aims to characterize the hoop stress and strain during the transients, as well as predict fission product migration and radiological release. Due to the startup conditions of the HTGR-TR, the silicon carbide (SiC) layer for both of the transients and steady-state conditions are initially in compression for both the hoop stress and strain. The DLOFC transient at the beginning of cycle was the most challenging overall to the fuel, having the closest to tensile behavior and the closest to tensile strain. Because the integrity of the SiC layer was not challenged during any of the transients, the release of 90Sr and 137Cs was found to be negligible. The PLOFC transient at the end of cycle resulted in the largest 110mAg release and subsequent radioactivity increase. However, the increase in radioactivity was minor when compared to typical operating conditions due to the relatively large diffusivity of 110mAg though SiC at normal operating temperatures. Highlights include Neutronics and thermal-hydraulic modeling generating steady-state and transient conditions for use in fuel performance analysis.2. Detailed BISON analysis demonstrating the excellent performance of TRISO fuel during PLOFC and DLOFC transients.3. Fuel behavior showing negligible fission product release.

Optimum power determination and comparison of multi-pass and OTTO fuel management schemes of HTGR-based PeLUIt reactor

A co-generation reactor based on pebble-bed High-Temperature Gas-Cooled Reactor (HTGR) has been designed. This study focuses on designing a modular nuclear reactor based on pebble bed high-temperature reactor (Electric Power and Industrial Hot Steam Plant/Pembangkit Listrik dan Uap Panas Industri, abbreviated as PeLUIt), an upgrade from Indonesia’s 10 MW Experimental Power Reactor (Reaktor Daya Eksperimen, RDE). PeLUIt is proposed as a solution for electricity and co-generation application. The 5-pass and Once-Through-Then-Out (OTTO) refueling scheme for the PeLUIt has been analyzed. PEBBED was used for equilibrium core calculations, whereas COMBINE7 was used for spectrum calculations utilizing ENDF/B-VII.0 as nuclear data. Based on the results, it can be inferred that the PeLUIt co-generation reactor analysis with an optimal power of 40 MWt is produced by increasing fuel intake into the core while retaining core geometry and fuel design. In this study, burn-up levels for each refueling scheme are also analyzed, and the value for OTTO is lower than 5-pass refueling schemes. In addition, the Depressurized Loss of Forced Cooling (DLOFC) accident scenario for both types of refueling schemes was also analyzed. The result showed that the OTTO refueling scheme has a higher safety level than the 5-pass refueling scheme.

Evaluation of the Basic Neutronics and Thermal-Hydraulics for the Safety Evaluation of the Advanced Micro Reactor (AMR)

South Africa requires safe affordable distributed base load energy, one way to achieve this is to use nuclear power integrated with renewable energy sources on a decentralized basis. This suggests the development of its own micro modular nuclear reactor, to supply energy to towns, small communities, mines and processing plants. Large Light Water Reactors (LWRs) are expensive and require a large infrastructure development. A High Temperature Reactor (HTR) called the Advanced Micro Reactor (AMR) is in the process of being developed and the design philosophy is to design for inherent safety, maximally using technology that has been developed and validated in previous HTR programs albeit in a completely different and unique configuration. The concept is based on existing knowhow and experience/expertise in South Africa during the time of the Pebble Bed Modular reactor (PBMR) project. These AMR reactors are to be factory built to obtain good quality control and rolled out to various sites. Once the reactor has reached its end of life, it would be returned to a licensed organisation for refuelling. The AMR produces 10MW of thermal power. The reactor configuration uses hexagonal graphite blocks for structural and moderator material, which are arranged to form a cylindrical core layout. The fuel assemblies are silicon carbide tubes that house coated particle fuel, immersed in a lead-bismuth eutectic alloy (LBE). Each fuel assembly is contained in a boring within the graphite moderator that allows an annulus for cooling. There are 420 fuel assemblies in the core. Low enriched fuel in the form of UO₂ or UCO is used. Helium gas is used as coolant. The coolant enters the core at 450°C and exits at 750°C. The mechanical, neutronic and thermal-hydraulic design of the AMR, is being evaluated with assistance from STL Nuclear (Pty) Ltd., the University of Pretoria (UP), the North-West University and the South African Nuclear Energy Corporation (NECSA). The OSCAR-5 code package, together with the Serpent neutronic code were used to perform the basic neutronic studies while the Flownex package was used to determine the thermal-hydraulic and safety evaluation for the Design Base Accident (DBA) specifically the Depressurized Loss of Forced Cooling (DLOFC) event.

A comprehensive Thermo-hydraulic neutronic and safety analysis of a 100MWth pebble bed reactor core

The HTMR100 is a 100MWth High Temperature Gas-cooled Reactor (HTGR), which is helium-cooled and graphite moderated. This reactor is of the Pebble Bed type featuring a Once-Through-Then-Out (OTTO) fuelling scheme, this cycle was chosen for the simplicity of the fuel handling system.The HTMR100 Pebble Bed Reactor (PBR) has a core volume of 27.73 m3 (core diameter 2.6 m, core height ∼ 5.2 m) which results in an average core power density of 3.6 MW/m3. This reactor utilizes uranium dioxide fuel (UO2) at a U-235 enrichment of 10 wt% and 10 g of heavy metal (total uranium) per Fuel Sphere (FS). Although there are various types of fuel options that can be utilized in the HTMR100 reactor, ranging from LEU to mixtures of Th/LEU, Th/HEU and Th/Pu, this study will only focus on LEU in the form of UO2 fuel.This study investigates the neutronic and thermal–hydraulic modelling of the 100MWth HTMR100 at equilibrium (100% power with the control rods situated at the nominal position for a core having an equilibrium fuel burnup profile) and at transient/accident conditions. The purpose of the investigation is to determine if the maximum fuel temperature remains below the set limit of 1600 °C during a design-base accident event that is defined as a Loss of Coolant Accident (LOCA).The equilibrium core was simulated using the Very Superior Old Programs (VSOP99) suite of codes to obtain burnup results. The transient behaviour of the core was modelled by a code called Multi-group Time Dependent Neutronics and Temperatures (MGT). MGT assesses the dynamic transient results of the accident scenarios such as the Depressurized Loss of Forced Cooling (DLOFC), Pressurized Loss of Forced Cooling (PLOFC), Control Rod Withdrawal (CRW) and a Control Rod Ejection (CRE) as well as normal operating transients of which a Load Following (LF) operation is a typical example.The HTMR100 reactor utilizing the enrichment and heavy metal loading as specified does indeed produce the targeted 80 000 MWD/THM burnup for the OTTO fuel cycle. The study also proves that the VSOP99 and the MGT codes do in fact yield similar results for the HTMR100 reactor with regards to fuel centreline temperatures, outer sphere surface temperatures as well as moderator temperatures for the postulated accident scenarios that were analysed. The results also indicate that the design-base transient safety analysis simulations prove that the fuel temperatures remain below the set value of 1600 °C for the oxide-based fuel. A temperature-volume analysis of a DLOFC design-base event shows that only ∼ 2 % of the fuel reaches the higher temperatures of 1500 °C to 1600 °C in a DLOFC event.The beyond-design base events with a very low probability of occurrence do however exceed the set value of 1600 °C but the fuel only exceeds this set limit for short periods of time while the transient occurs. The probability that this will lead to fuel damage should be very small since the time at these high temperatures is short and the volume fraction of the core exposed to these high temperatures is small.

Transient coupling analysis of loss-of-coolant accidents in HTR-PM based on CFD approaches

Pressurized loss of forced cooling (PLOFC) and depressurized loss of forced cooling (DLOFC) are two important accidents in high temperature gas-cooled reactors. In the present study, it is intended to investigate thermal–hydraulic behaviors of the reactor core under these two accidents using the CFD approach. To this end, a coupling solver is developed based on the OpenFOAM platform. The computational domain is a 12° wedge-shaped three-dimensional region for the core of HTR-PM. The performed analyses demonstrate that the maximum relative error is less than 3.9%, indicating the accuracy and reliability of the established model. Moreover, it is found that there exists a short period when the heat is transferred from the coolant helium to the pebble-bed during the initial stage of PLOFC case. The maximal fuel temperature of 1195 ℃ occurs at t = 15.3 h for PLOFC, and the maximal fuel temperature of DLOFC is 1449 ℃ at t = 26 h. The obtained results reveal that sufficient safety margin is ensured under the PLOFC and DLOFC accident scenario.

Optimization of PeLUIt-150 Conceptual Design

A 150MWt pebble bed high temperature reactor, called PeLUIt-150, is proposed as a solution for distributed energy demand including the electricity and various co-generation potentials. In current study an optimization of the design is performed. The optimization includes its optimum power density to have a compact core design including the height-to-diameter (H/d) ratio. As several computer codes is used in this study, PEBBED-THERMIX codes, comparison of results from different code is also discussed. Constrain of the design at current stage are the maximum fuel temperature under depressurized loss of forced cooling (DLOFC) accident which represent the passive performance of the core, and discharge burnup which represent the fuel utilization performance of the core. Current study achieves an optimized design with core diameter and height of 260 cm and 601.11cm, a total active core volume of 31.9 m3. The average power density is 4.7W/c m3 with discharge burnup of 80.82MWd/Kg and the maximum fuel temperature after DLOFC is 1619.9°C.

Pebble Bed High Temperature Reactor for Electric and Cogeneration Application

Specific energy demand of Indonesia with its distributed energy demand required a special energy supply solution. The energy supply solution should be able to fulfil the electric demand of huge and growing population which spread in many different islands also to fulfil the energy demand to process abundances yet distributed natural sources. This paper proposes a pebble bed high temperature reactor as an energy supply in the form electric and/or cogeneration application. A parametric survey of the core geometry of a 150MWt Pebble Bed Reactor called PeLUIt is performed, follow by an equilibrium analysis of the optimized design. An accident analysis of depressurized loss of forced cooling (DLOFC) is performed to show the strong passive safety features of the design. Finally the heat-mass balance of the plant is also presented.

The development of TRIAC-BATAN: a triso fuel performance analysis code

A computer code for analyzing the safety feature of triso coated particles for high temperature reactor has been developed based on PANAMA code. This python based code were separated into several modules to perform the main TRIAC-BATAN and linear interpolation calculation, to read the input data file, and to control the sequence of all TRIAC-BATAN calculation. In this initial development phase, TRIAC-BATAN can produce similar pattern as PANAMA calculation in the case of failure fraction of triso particles, either for Depressurized Loss Of Forced Cooling (DLOFC), 100°C higher than DLOFC temperature and constant accident temperature at 1600°C. By using the euclidean distance, TRIAC-BATAN calculation and PANAMA for DLOFC condition are separated in around 3.23 ⋅ 10−7.

Equilibrium Core Design of Reaktor Daya Eksperimental

A prototype of a high temperature gas-cooled reactor for electric and cogeneration application, called Reaktor Daya Eksperimental (RDE) is being developed at BATAN. RDE is developed to fulfill the huge yet distributed Indonesian energy demand, in particular for the eastern part of Indonesia. As a moving fuel core, analysis of the fuel management of PBR can be divided into the analysis of the start-up phase where the core composition, neutron flux, and power density profile still changing and equilibrium core. Equilibrium core design usually represent the general performance over the entire range of reactor operation. As part of the design development, current paper presents the equilibrium core design of RDE and depressurized-loss-of-forced-cooling (DLOFC) accident analysis. PEBBED code is utilized for the equilibrium core analysis. Results of this study include several possible equilibrium designs for RDE which then the design with 123 fuel pebble recirculated per day with equilibrium keff 1.02571, discharge burnup of 79.36 and maximum power generation 0.66kW per pebble fuel is chosen. Maximum fuel temperature of this design under the DLOFC accident is 1015.9°C which occurred 8.5h after the initiation of the accident. Results of DLOFC accident shows the strong passive safety features of RDE.

Study on the DLOFC and PLOFC accidents of the 200 MWe pebble-bed modular high temperature gas-cooled reactor with TINTE and SPECTRA codes

Validation and verification (V&V) is important and receives significant attention during the design and analysis of the nuclear power plant. As the first commercial modular high temperature gas-cooled reactor (HTR) demonstration plant, during the regulatory review of the preliminary safety analysis report (PSAR) and the final safety analysis report (FSAR) of the 200 MWe Pebble-bed Modular High Temperature gas-cooled Reactor (HTR-PM), the authorities also pay high attention to the code validation and verification.Within the cooperation between the Nuclear Research Group (NRG) of Netherland and Institute of Nuclear and new Energy Technology (INET), Tsinghua University of China, a code to code verification was carried out between the TINTE code, a thermal-hydraulic design and accident analysis tool for the Pebble-bed High Temperature Gas-cooled Reactor (HTGR), and the SPECTRA code, a thermal-hydraulic analysis code developed at the NRG. Three typical accidents scenarios, two depressurized loss of forced cooling (DLOFC) accidents scenarios and one pressurized loss of forced cooling (PLOFC) accident scenario, have been analyzed both with the TINTE code and the SPECTRA code.The simulation results show good agreement between TINTE code and SPECTRA code. The results also indicate that, due to the inherent safety feature of the HTR-PM design, the maximal fuel temperature during above accidents would never exceed the limitation of 1620 °C, below which there is no additional damage for the TRISO particles, and the SiC layer of each particle can guarantee the fission-product-retention capacity. This work will be helpful to the further study of the HTGR, and more code to code verification is planned.

Numerical analysis of 1/28 scaled NGNP HTGR reactor building test facility response to depressurization event using GOTHIC

A depressurized loss of forced cooling (DLOFC) accident is an important type of accident scenario in high temperature gas reactor (HTGR) design, which is initiated by a break in the helium pressure boundary (HPB). This class of accident scenarios results in a depressurization of the primary helium coolant system with the subsequent release of helium into the reactor building (RB) and the atmosphere through vented low pressure containment (VLPC). After the total depressurization of helium, depending on the specific accident scenarios, it is also possible that air enters into the reactor pressure vessel (RPV) through the RB, which can potentially react with fuel and other nuclear-grade graphite components.In this study, generation of thermal hydraulic information in containments (GOTHIC) model of a 1/28-scaled simplified test facility was developed to analyze the depressurization scenarios and validate them against experimental data. Natural leakage from the test facility was modeled using two different methods. The post-depressurization refill of air into the test facility compartments along with two hypothetical depressurization scenarios were analyzed and compared with experimental data. In general, simulation results were observed to be consistent with the experimental data.

Preliminary phenomena identification and ranking tables on the subject of the High Temperature Gas-cooled Reactor-Pebble Bed Module thermal fluids and accident analysis

The High Temperature Gas-cooled Reactor-Pebble Bed Module (HTR-PM) is a demonstration power plant designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University of China. In conformance with China’s nuclear safety requirements related to the development, assessment, application and quality assurance of computer software used in the safety analysis of nuclear power plants (NPPs), a phenomena identification and ranking table (PIRT) process was conducted on the subject of the HTR-PM thermal fluids and accident analysis by an expert panel sponsored by INET, on the basis of the latest HTR-PM design scheme. The PIRT objective is to identify safety-relevant phenomena that may occur during normal operation and under accident conditions, and to assess research activities pertaining to code development and assessment, such as the code validation in INET.In the PIRT process, safety-related phenomena were identified by the expert panel for the following HTR-PM scenarios: normal operation (NO), pressurized loss of forced cooling (PLOFC), depressurized loss of forced cooling (DLOFC), air ingress following DLOFC, steam/water ingress and anticipated transients without scram (ATWS). Subsequently, the relative importance of each phenomenon was evaluated according to its effect or influence on the figures of merit established by the expert panel, such as the fuel failure fraction and the maximum fuel temperature. In addition, each identified phenomenon was assigned a knowledge level rank by the expert panel, indicating the prediction capability for this phenomenon based on existing analytical tools and test data. What is more, all of the PIRT elements were well documented by a special report, including PIRT issues, objectives, descriptions of the HTR-PM plant and accident scenarios, evaluation criteria, current knowledge base, ranking scales of relative importance and knowledge level, a set of PIRTs, and expert discussions.

Neutron poison distribution in the central reflector to reduce the DLOFC temperature of a Th-LEU fueled OTTO PBMR DPP-400 core

ABSTRACTIn a previous study using a mixture of thorium and 20 a/o% LEU at 16 gram per fuel sphere heavy metal loading and adjusting the effective fuel enrichment to produce the same amount of cumulative energy per fuel sphere as with the 10 a/o% Low Enriched Uranium (LEU), the maximum Depressurized Loss Of Forced Cooling (DLOFC) temperature was reduced from 2273 to 1925 °C and 1811 °C for a symmetric and asymmetric core, respectively using an once-through-then-out (OTTO) fuelling scheme. This article presents an additional strategy for reducing the maximum DLOFC temperature by placing an optimized distribution of neutron poisons in the central reflector. This strategy produced maximum DLOFC temperatures of 1509 and 1448 °C for the symmetric and the asymmetric cores, respectively. These results are impressive as it means that the less complicated OTTO cycle with its lower capital cost achieved the same cumulative energy produced per fuel sphere than the standard six-pass refuelling scheme and that at substantially lower maximum DLOFC temperatures. Both the addition of the neutron poisons to the central reflector and the creation of a radially asymmetric core resulted in lower burn-ups that had to be reversed by increasing the enrichment of the fuel.

Experimental study on 1/28 scaled NGNP HTGR reactor building test facility response to depressurization event

The analysis and understanding of air ingress events are an important aspect of the design of high-temperature gas-cooled reactor (HTGR) accident scenarios. These include depressurized loss of forced cooling (D-LOFC) events that allow for the possibility of air ingress into the reactor pressure vessel as a result of a break in the helium pressure boundary, which can ultimately result in oxidation of the fuel elements and other nuclear-grade graphite components. To characterize air ingress into the vented low-pressure containment of the next generation nuclear plant HTGR during hypothetical moderate-sized D-LOFC break accidents experimentally, a 1/28 scaled simplified reactor building model was established. A non-dimensional similarity approach was employed for scaling of the experimental facility. Three experiments were designed and conducted to study the dynamic response to the accident scenario. The experimental results suggest an increment in the flow area of the check valve between the reactor cavity (CV1) and steam generator cavity (CV3). Furthermore, qualitative analysis was conducted on the experimental data.

Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

The present work investigates the running-in phase of a 100MWth Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated.For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme.Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7years after start of reactor operation.The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the equilibrium core during every stage of the running-in phase. The maximum fuel temperature during a Depressurized Loss of Forced Cooling (DLOFC) with scram remains below 1400°C throughout the running-in phase, quite a bit below the TRISO failure temperature of 1600°C. The uniform reactivity coefficients of cores with U-235/U-238 driver fuel are much stronger negative compared to U-233/Th driver fuel, which implies that the stronger reactivity insertion by water ingress and the reactivity addition by xenon decay during a DLOFC without scram can be compensated without fuel temperatures exceeding 1600°C.

Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30g thorium pebbles, for enhanced conversion.The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600°C.Results show that high conversion ratios (CR >0.96) and passive safety can be combined in a thorium PBR within a practical operating regime, which means a thermal power of 100MW or higher, 1000days total residence time of the breeder pebbles and fuel pebble handling times longer than 14.5s, like in the HTR-PM. With an increased U-233 content of the fresh driver pebbles (18w%), breeding (CR=1.0135) can already be achieved for a 220cm core and 80cm driver zone radius. While the decay heat removal is sufficient in this design, the temperature feedback of the undermoderated driver pebbles is too weak to compensate the reactivity insertion due to the xenon decay during a DLOFC without scram. With a lower U-233 content per driver pebble (10w%) it was found possible to combine breeding(CR=1.0036) and passive safety for a 300cm core and 100cm driver zone radius, but this does require more than a doubling of the pebble handling speed and a high reprocessing rate of the fuel pebbles. The maximum fuel temperature during a DLOFC without scram was simulated to be 1481°C for this design, still quite a bit below the TRISO failure temperature. The maximum reactivity insertion due to an ingress of water vapour is also limited with a value of +1497pcm.

Uncertainty and Sensitivity Analyses of a Pebble Bed HTGR Loss of Cooling Event

The Very High Temperature Reactor Methods Development group at the Idaho National Laboratory identified the need for a defensible and systematic uncertainty and sensitivity approach in 2009. This paper summarizes the results of an uncertainty and sensitivity quantification investigation performed with the SUSA code, utilizing the International Atomic Energy Agency CRP 5 Pebble Bed Modular Reactor benchmark and the INL code suite PEBBED-THERMIX. Eight model input parameters were selected for inclusion in this study, and after the input parameters variations and probability density functions were specified, a total of 800 steady state and depressurized loss of forced cooling (DLOFC) transient PEBBED-THERMIX calculations were performed. The six data sets were statistically analyzed to determine the 5% and 95% DLOFC peak fuel temperature tolerance intervals with 95% confidence levels. It was found that the uncertainties in the decay heat and graphite thermal conductivities were the most significant contributors to the propagated DLOFC peak fuel temperature uncertainty. No significant differences were observed between the results of Simple Random Sampling (SRS) or Latin Hypercube Sampling (LHS) data sets, and use of uniform or normal input parameter distributions also did not lead to any significant differences between these data sets.

Thermal-hydraulic design and transient evaluation of a small long-life HTR

Small long-life high temperature gas-cooled reactors (HTRs) may provide electricity or heat for remote areas or industrial users in developed and/or developing countries. Moreover, small HTRs have advantages over large nuclear reactors of demonstrated inherent safety, transportability, modular construction, and flexible site selection. This paper presents the thermal-hydraulic evaluations of the U-Battery, which is a small, long-life and block-type HTR using the DALTON/THERMIX code system. The thermal-hydraulic characteristics of a cylindrical design and an annular design of the U-Battery were evaluated for loss of forced-cooling (LOFC) incidents including depressurized LOFC (DLOFC) and pressurized LOFC (PLOFC) incidents. The calculations show that the stronger natural circulation during the PLOFC makes the reactor core cool faster than during the DLOFC, flattens the radial solid temperature distribution, and transfers more heat from the hot regions (bottom and center of the reactor core) to cold regions (top and periphery of the reactor core). Although the natural circulation in the reactor core is so weak that it is neglected during the DLOFC, the decay heat is removed passively by conduction without any violation of the temperature limits for the 20MWth U-Battery. The comparisons of the cylindrical and annular reactor core configurations show that the latter is a better design with a lower maximum core temperature during the LOFC because of the central graphite reflector. Moreover, it is possible to adopt the current reactor configuration when the thermal power of the U-Battery increases from 20MWth to 40MWth.

Preliminary reactor physics assessment of the HTR module with 14% enriched UCO fuel

The high temperature reactor (HTR) Module (Lohnert, 1990) is a graphite-moderated, helium cooled pebble bed design that has been extensively used as a reference template for the former South African (Matzner, 2004) and current Chinese (Zhang et al., 2009) HTR programs. This design utilizes spherical fuel elements packed into a dynamic pebble bed, consisting of tri-structural isotropic (TRISO) coated uranium oxide (UO2) 500μm fuel kernels with a U-235 enrichment of 7.8% and a heavy metal loading of 7g per pebble. This fuel type was previously qualified for use in Germany for pebble bed HTRs, as well as undergoing re-qualification in South Africa for the PBMR project. It is also the fuel type being tested for use in the high temperature reactor (HTR-PM) under construction in China.In the United States, however, a different TRISO fuel form is the subject of a qualification program. The U.S. experience with HTRs has been focused upon the batch-fueled prismatic reactor in which TRISO particles are embedded in cylindrical compacts and stacked inside the graphite blocks which comprise the core. Under this type of operating regime, a smaller TRISO with a different composition and enrichment performs better than the fuel historically used in PBRs. Fuel kernels and compacting techniques more suited to prismatic core duty are currently being developed and qualified under the U.S. Department of Energy's Advanced Gas Reactor (AGR) fuel development program and in support of the Next Generation Nuclear Plant project.Interest in the pebble bed concept remains high, however, and a study was undertaken by the authors to assess the viability of using AGR fuel in a pebble bed reactor. Using the German HTR Module as the reference plant, key neutronic and thermal-hydraulic parameters were compared between the nominal design and one fueled with the fuel that is the focus of the AGR program.Seven cases, varying the total heavy metal loading and number of fuel passes, were analyzed, and recommendations are made for achieving a feasible UCO-fueled HTR Module design from a reactor physics point of view. The PEBBED-THERMIX code (Gougar et al., 2010b), developed specifically for the analysis of pebble bed HTRs, was used to compare the coupled neutronic and thermal fluid performance of the two fuel particle designs. For seven variations on the fuel loading scheme, comparisons were made of steady state core multiplication factor (k-eff) at a specified burnup, control rod worths, power and flux profiles, fuel and moderator temperatures, and power peaking factors. Also analyzed were the maximum fuel temperatures during a depressurized loss of forced cooling (DLOFC) event, as well as the reactivity behavior for a water/steam ingress scenario.The first three cases compared the HTR-Module fuel (Case 1) with fuel of the same UO2 TRISO particle design with enrichment increased to 14% (Case 2) and UCO TRISO fuel as described above. For these three cases the sphere heavy metal loading and average number of passes through the core remained constant at 7g and 15 passes. The analysis of the normal operation (steady state) equilibrium results for these three cases showed that the dominant contributor to the observed variances between the HTR Module UO2 and UCO cores is the increase in the U-235 enrichment to 14%, and not the additional moderation effects of the oxygen to carbon exchange.In addition to the steady-state analyses, two important design basis accidents were also included in this study. The first was an extreme loss of forced cooling accident caused by a large double-ended guillotine break. This event is the limiting case for the fuel temperatures. The second event was the ingress of water (in the form of steam) into the core, because of a hypothetical steam generator tube rupture. The use of UCO fuel at 14% enriched (Case 3) lead to a small increase of 48°C in the DLOFC peak fuel temperature to 1533°C. If typical uncertainty margins between 4% and 7% are taken into account, the maximum fuel temperatures are below 1660°C. However, it was also shown that only 4% of the 360,000 fuel spheres in the core have maximum temperatures above 1400°C. It was also confirmed that the DLOFC fuel temperature data for Cases 2 and 3 are essentially identical, confirming that the change in enrichment is the dominant driving factor in the differences observed between Cases 1 and 3.It was also found that the 7g heavy metal UCO fueled core is more reactive than the 7g heavy metal UO2 fueled core for a steam ingress event (1.29% for Case 1 and 1.56% for Case 3 at 660kg steam). However, the control rod shutdown worths for a full SCRAM were also compared for these two cases (5.16% and 4.30%, respectively), and it was concluded that an acceptable shutdown margin exists for both cases. It therefore seems feasible, from a water ingress point of view, to operate an HTR Module core design with UCO fuel enriched to 14% and loaded to 7g heavy metal. By comparing the results for Cases 1–3, it was shown that the higher enrichment plays the dominant role in the UCO fueled core's reactivity behavior, and not the change to UCO fuel kernels.The analysis of these two accidents, together with the acceptable results obtained from the steady-state and the control rod worth analysis, provided a preliminary insight into the behavior of the UCO-fueled HTR Module design. It is important to note that a significant analysis gap still exist in evaluating the control rod worths and water ingress effects at other core conditions, for example hot and cold shutdown conditions. A final statement on the feasibility of this fuel can only be made at that point.The fuel performance of this core design was beyond the scope of this study, and care should be taken not to equate acceptable neutronics and thermal fluid behavior with acceptable levels of fission product release rates, since many more factors are involved in this aspect of fuel design. Nonetheless, this preliminary analysis suggests that the fuel being qualified by DOE for deployment in prismatic HTRs can also be used in pebble bed reactors.