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This presentation covers the work performed quantifying the effects of stochastic pebble and TRISO packing a the HTR-Proteus pebble-bed reactor experiment.
The HTR-10 was used as a representative pebble-bed high-temperature gas-cooled reactor in this assessment of nuclear data’s impact on important reactor and spent fuel metrics, including safety-related quantities such as the effective multiplication factor (k eff ), temperature reactivity feedback, spent fuel inventory, and decay heat. Using the SCALE code system tools and ENDF/B-VII.1 nuclear data libraries, we quantify the effect of nuclear data uncertainties on these key performance metrics for both fresh fuel and equilibrium core configurations. For reactor core key parameters, important contributors to uncertainty include reactions of 235 U [$\bar{v}$, fission, (n, γ)], 238 U [elastic, (n, γ)], and graphite [elastic, (n, γ)]. Additional important contributors for the equilibrium core include reactions of higher actinides ( 239 Pu, 240 Pu) and fission products ( 135 Xe, 149 Sm). For spent fuel analysis, most nuclide inventory uncertainties remain below 5%. Higher uncertainties up to 11% are being observed for minor actinides like 243 Am and 244 Cm. Additionally, fission product uncertainties in 155 Eu and 155 Gd, of 25% and 23% respectively, are also significant and have implications for burnup credit applications. 110m Ag also shows high uncertainty of up to 11%, mainly due to fission product yield uncertainties. Decay heat relative uncertainties remain below 0.6% up to 10 years’ cooling time after fuel discharge. The highest relative uncertainty of 1.5% occurs at 500 years of cooling; however, because the decay heat value is very low at that time, the absolute uncertainty is not significant. This work demonstrates that extending assessments beyond fresh fuel k eff to include irradiated cores, nuclide inventories, and decay heat is essential in understanding the behavior of uncertainties as a function of fuel burnup and can support improvements of safety margins and spent fuel management.
China is now designing and constructing a high temperature gas cooled reactor-pebble bed module (HTR-PM). In order to investigate the future decommissioning approach and evaluate possible radiation dose, gamma dose rate near the reactor pressure vessel was calculated for different cooling durations using QAD-CGA program. The source term of this calculation was provided by KORIGEN program. Based on the calculated results, the spatial distribution and temporal changes of gamma dose rate near reactor pressure vessel was systematically analyzed. A suggestion on planning decommissioning operation of reactor pressure vessel of HTRPM was given based on calculated dose rate and the Chinese Standard GB18871-2002. (authors)
Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the ²³⁵U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.
Information and measured data from the intial Fort St. Vrain (FSV) high temperature gas reactor core is used to develop a benchmark configuration to validate computational methods for analysis of a full-core, commercial HTR configuration. Large uncertainties in the geometry and composition data for the FSV fuel and core are identified, including: (1) the relative numbers of fuel particles for the four particle types, (2) the distribution of fuel kernel diameters for the four particle types, (3) the Th:U ratio in the initial FSV core, (4) and the buffer thickness for the fissile and fertile particles. Sensitivity studies were performed to assess each of these uncertainties. A number of methods were developed to assist in these studies, including: (1) the automation of MCNP5 input files for FSV using Python scripts, (2) a simple method to verify isotopic loadings in MCNP5 input files, (3) an automated procedure to conduct a coupled MCNP5-RELAP5 analysis for a full-core FSV configuration with thermal-hydraulic feedback, and (4) a methodology for sampling kernel diameters from arbitrary power law and Gaussian PDFs that preserved fuel loading and packing factor constraints. A reference FSV fuel configuration was developed based on having a single diameter kernel for each of the four particle types, preserving known uranium and thorium loadings and packing factor (58%). Three fuel models were developed, based on representing the fuel as a mixture of kernels with two diameters, four diameters, or a continuous range of diameters. The fuel particles were put into a fuel compact using either a lattice-bsed approach or a stochastic packing methodology from RPI, and simulated with MCNP5. The results of the sensitivity studies indicated that the uncertainties in the relative numbers and sizes of fissile and fertile kernels were not important nor were the distributions of kernel diameters within their diameter ranges. The uncertainty in the Th:U ratio in the intial FSV core was found to be important with a crude study. The uncertainty in the TRISO buffer thickness was estimated to be unimportant but the study was not conclusive. FSV fuel compacts and a regular FSV fuel element were analyzed with MCNP5 and compared with predictions using a modified version of HELIOS that is capable of analyzing TRISO fuel configurations. The HELIOS analyses were performed by SSP. The eigenvalue discrepancies between HELIOS and MCNP5 are currently on the order of 1% but these are still being evaluated. Full-core FSV configurations were developed for two initial critical configurations - a cold, clean critical loading and a critical configuration at 70% power. MCNP5 predictions are compared to experimental data and the results are mixed. Analyses were also done for the pulsed neutron experiments that were conducted by GA for the initial FSV core. MCNP5 was used to model these experiments and reasonable agreement with measured results has been observed.
Abstract – The Gas-Cooled Reactor deployment effort in the United States is a collaboration between the US government and private vendors. The US Department of Energy sponsors the qualification of fuels, materials, and analysis methods that exploit the capabilities of the national laboratories (mainly Idaho, Oak Ridge, and Argonne) to generate the fundamental fuel, material, and core behavior data needed to support design and licensing efforts by vendors in general. Private HTR vendors provide input to the R&D program and engage in the design of the systems, structures, and components that are unique to their design concepts. Their plant design all exploit the properties and behavior of the TRISO fuel, graphite, and alloys being tested in the DOE qualification programs. This paper will summarize the goals and recent progress of the US Gas-Cooled Reactor Program.
The behavior of the graphite dust is important to the safety of High Temperature Gas-cooled Reactors. This study investigated the flow of graphite dust in helium mainstream. The analysis of the stresses acting on the graphite dust indicated that gas drag played the absolute leading role. Based on the understanding of the importance of gas drag, an experimental system is set up for the research of dust releasing behavior in accident. Air driven by centrifugal fan is used as the working fluid instead of helium because helium is expensive, easy to leak which make it difficult to seal. The graphite particles, with the size distribution same as in HTR, are added to the experiment loop. The graphite dust releasing behavior at the loss-of-coolant accident will be investigated by a sonic nozzle. (authors)
The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] it is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)
The paper will describe the design concept of the emergency power system utilized in the high temperature gas cooled reactor project (HTR-PM) which is currently under construction in Shandong, China. As a class 1E system which is related to the safety feature of the reactor, emergency power system is responsible for providing multiple reliable power sources for nuclear safety class loads. Therefore, the structure, power source deployment, operation, maintenance and alignment of the system should meet several standards and regulations. The paper will give a detailed description about the designing process of the system which include structure design, determination of primary device parameters, and some calculation results of device capacity such as emergency diesel engine, UPS et al. (authors)
Through multi-modular scheme, i.e. the superheated steam flows produced by multiple MHTGR-based nuclear supplying system (NSSS) modules are combined together to drive a common thermal load, the inherent safety feature of MHTGR can be applicable to the large-scale nuclear plants at any desired power ratings. Since the plant power control technique of classical single-modular plants cannot be directly applied to multi-modular plants, it is necessary to develop the power control method of the multi-modular plants. In this paper, the plant power control strategy of the two-modular HTR-PM plant, i.e. the first under-constructed multi-modular high temperature gas-cooled nuclear plant in the world is given, which includes both the feedback loops and control algorithms. Then, the control strategy is verified through numerical simulation, and the simulation results in the cases of plant power step and ramp show the satisfactory performance of this control strategy. (authors)
The helium circulator is one of the key equipment in the first loop of the High Temperature gas-cooled Reactor Pebble bed Module (HTR-PM). The active magnetic bearings are employed in the helium circulator, so that the inlet ring sealing clearance is about 5 to 10 times larger than that if the mechanical bearings are employed. Bigger clearance results in bigger leakage flow and bigger axial force. A new design of balance holes was proposed to reduce the axial aerodynamic force and the design principle was analyzed. To investigate the axial aerodynamic force due to leakage flow, numerical simulation of fluid dynamic was conducted in the impeller with two kinds of inlet ring sealing clearance and four types of balance holes, respectively. The calculation results shows that the balance holes with diameter 12 mm have high performance and that the designed balance holes reduce effectively about 78% of the axial aerodynamic force. The axial force was obtained from the measured real-time electric current of the axial active magnetic bearings in one helium circulator prototype. The comparison of experimental data with the CFD analysis results showed that the new design of balance holes was successful. (authors)
In the fifth, sixth, and seventh Framework Program of the European Commission, the Nuclear Research and Consultancy Group (NRG) have conducted a large graphite irradiation program for the development of High Temperature Reactor technology. The graphite irradiation program comprises five irradiations executed in the High Flux Reactor (HFR) in Petten, resulting in design curves at irradiation temperatures between 650 deg. C and 950 deg. C. The neutron doses of these irradiations, called INNOGRAPH, range from 1 dpa to 23 dpa. The irradiation facilities contained present day available graphite grades: NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, LPEB, IG-110 and IG-430. The post-irradiation examination of the INNOGRAPH samples is focused on dimensional changes, dynamic Young s modulus, coefficient of thermal expansion (CTE) and coefficient of thermal conductivity. Irradiation behaviour of HTR candidate graphite grades are compared with respect to manufacturing process and raw coke material. Accuracy and reliability of irradiation parameters and post irradiation examination methods are essential for safety and performance relevant material property determination, in particular for probabilistic material behaviour such as graphite. Considerations on quality requirements with regards to graphite material property determination are therefore included in the paper. (authors)
Active magnetic bearing (AMB) has been widely used in high-temperature gas-cooled reactor (HTR) for its specific properties: free of contact, absence of lubrication, capability of rotating at high speed. AMB system is inherently unstable since positive feedback exists in the open loop system and, thus, requires feedback control. From the control view of point, accurate system model is essential for designing of an effective controller especially when the flexible gyroscopic rotor is involved. In the paper, a flexible rotor and an open-loop AMB system model were established for the purpose of controller design. Dynamic characteristics of the undamped flexible rotor are analyzed including modal vibration analysis and undamped natural frequencies. Additionally, some characteristics of open-loop AMB system are investigated such as relationship between support stiffness and natural frequencies, and the system frequency response. The analysis results in this paper will provide valuable information for AMB system feedback controller design. (authors)
This assessment explores the current analysis methodologies of various external and MOOSE-based tools available at the Idaho National Laboratory to model pebble bed reactor multiphysics phenomena. The study involves three MOOSE-based, coupled codes: 1) MAMMOTH for reactor physics, 2) Rattlesnake for neutron transport and 3) Pronghorn for thermal fluids and strongly coupled conjugate heat transfer. These advanced tools are being developed under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program (NEAMS). They currently contain a collection of advanced modeling capabilities that allow high-fidelity simulations of advanced reactor concepts including gas and fluoride-salt cooled pebble bed reactors. Software development of these tools follows NQA-1 standards ensuring software quality that many of the currently deployed legacy codes lack. In this work, the MAMMOTH Pronghorn tools are compared to the legacy codes PEBBED-THERMIX-KONVEK. The external codes in use are LAMMPS for pebble mechanics and Serpent for cross section preparation. The coupling under MOOSE enables significant flexibility for modern engineering tools since MOOSE applications can be seamlessly coupled with various levels of resolution. This report focuses on low-resolution (r-z) and midresolution (3-D) homogenized geometries. The high-resolution Pebble Tracking Transport solution is not pursued herein and is postponed to future work. The results from this assessment show that the new MOOSE-based tools compare well with the results from legacy tools (r-z) and produce reasonable accuracy limited by the methods in the legacy tools. In the 3-D studies, the PJFNK-SPH methodology available in MAMMOTH allows the reproduction of Monte Carlo results with a transport-corrected diffusion solution. This has implications for the coupled solutions since without the SPH correction the temperature coefficient of reactivity is under-predicted and leads to lower temperature feedback and a significantly higher eigenvalue showing a difference of 2000 pcm.
As part of the program to demonstrate tristructural isotropic (TRISO)-coated fuel for the Next Generation Nuclear Plant (NGNP), Advanced Gas Reactor (AGR) fuel is being irradiation tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). This testing has led to improved kernel fabrication techniques, the formation of TRISO fuel particles, and upgrades to the overcoating, compaction, and heat treatment processes. Combined, these improvements provide a fuel manufacturing process that meets the stringent requirements associated with testing in the AGR experimentation program. Researchers at Idaho National Laboratory (INL) are working in conjunction with a team from Babcock and Wilcox (B&W) and Oak Ridge National Laboratory (ORNL) to (a) improve the quality of uranium oxycarbide (UCO) fuel kernels, (b) deposit TRISO layers to produce a fuel that meets or exceeds the standard developed by German researches in the 1980s, and (c) develop a process to overcoat TRISO particles with the same matrix material, but applies it with water using equipment previously and successfully employed in the pharmaceutical industry. A primary goal of this work is to simplify the process, making it more robust and repeatable while relying less on operator technique than prior overcoating efforts. A secondary goal is to improve first-pass yields to greater than 95% through the use of established technology and equipment. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 to November 2009. The AGR-1 fuel was designed to closely replicate many of the properties of German TRISO-coated particles, thought to be important for good fuel performance. No release of gaseous fission product, indicative of particle coating failure, was detected in the nearly 3-year irradiation to a peak burn up of 19.6% at a time-average temperature of 1038–1121°C. Before fabricating AGR-2 fuel, each fabrication process was improved and changed. Changes to the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a 6-inch diameter coater using a charge size about 21-times that of the 2-inch diameter coater used to coat AGR-1 particles. The compacting process was changed to increase matrix density and throughput by increasing the temperature and pressure of pressing and using a different type of press. AGR-2 fuel began irradiation in the ATR in late spring 2010.
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Pebble-Bed High-Temperature Gas-cooled Reactors (PB-HTGR) are moderated by the graphite in the fuel pebbles and the graphite reflector surrounding the pebble-bed. Because graphite is by far the most abundant material in PB-HTGRs and the primary moderator, accurate modeling of the graphite material, including density, impurities, and temperatures, is crucial for accurate computational modeling and simulation of these reactors. While main characteristics of the graphite components are often known, the local temperature is less well known and often averaged over all components. Here, this work studies the impact of considering accurate temperature profiles in the graphite material on the generation of a small PB-HTGR model at the state of equilibrium operation and on short-term accident progression. The fuel compositions for the PB-HTGR were determined using a jump-in equilibrium modeling method, the Axial Radial Zone Equilibrium Modeling (AR-ZEM) method. In contrast to previous work, the AR-ZEM method was used considering thermal-hydraulic feedback from the MELCOR code to determine temperatures of the fuel pebbles and the surrounding graphite reflector. The consideration of an axial and radial temperature profile in the core and reflector, as opposed to uniform material temperatures, had an impact of almost 1,300 pcm on the equilibrium core eigenvalue and caused significant differences in the discharged plutonium fuel inventory with up to 4.9% and 11.0% for Pu-239 and Pu-242, respectively. To assess the impact on short-term accident progression, two Anticipated Transient Without SCRAM (ATWS) events, a Pressurized Loss of Forced Coolant (PLOFC) and a Control Rod Withdrawal (CRW) with loss of flow, were simulated with MELCOR. The use of temperature profiles in the equilibrium core models did not reveal a significant impact on the temperature, power, or reactivity responses during the transients. In conclusion, a need for consideration of accurate temperature profiles, in particular for the graphite reflector, was found for the generation of equilibrium PB-HTGRs core models using jump-in methods, but detailed temperature profiles may not be necessary when performing conservative transient analysis.