Active Chemistry Control for Coolant Helium Applying High-Temperature Gas-Cooled Reactors

Lifetime extension of high-temperature equipment such as the intermediate heat exchanger of high-temperature gas-cooled reactors (HTGRs) is important from the economical point of view. Since the replacing cost will cause the increasing of the running cost, it is important to reduce replacing times of the high-cost primary equipment during assumed reactor lifetime. In the past, helium chemistry has been controlled by the passive chemistry control technology in which chemical impurity in the coolant helium is removed as low concentration as possible, as does Japan’s HTTR. Although the lifetime of high-temperature equipment almost depends upon the chemistry conditions in the coolant helium, it is necessary to establish an active chemistry control technology to maintain adequate chemical conditions. In this study, carbon deposition which could occur at the surface of the heat transfer tubes of the intermediate heat exchanger and decarburization of the high-temperature material of Hastelloy XR used at the heat transfer tubes were evaluated by referring the actual chemistry data obtained by the HTTR. The chemical equilibrium study contributed to clarify the algorism of the chemistry behaviours to be controlled. The created algorism is planned to be added to the instrumentation system of the helium purification systems. In addition, the chemical composition to be maintained during the reactor operation was proposed by evaluating not only core graphite oxidation but also carbon deposition and decarburization. It was identified when the chemical composition could not keep adequately, injection of 10 ppm carbon monoxide could effectively control the chemical composition to the designated stable area where the high-temperature materials could keep their structural integrity beyond the assumed duration. The proposed active chemistry control technology is expected to contribute economically to the purification systems of the future very high-temperature reactors.

A Study on an Evaluation of Indirect Cycle Plant Systems for HTGR

In the world now, several types of indirect system concept have been investigated for the High Temperature Gas cooled Reactor power plant (HTGR). From a point of optimization of HTGR, it is important to investigate and to compare their power conversion systems from a technical and an economical view point. In the first step of this study, an indirect steam cycle (ID-SC), an indirect gas turbine cycle (ID-GT), an indirect gas turbine combined cycle (ID-CCGT) and a direct gas turbine cycle (D-GT) has been chosen as the systems to be compared. The followings are chosen items for comparison analysis: a) Plant efficiency; b) Amount of commodities (which can estimate capital cost); c) Flexibility of reactor core design; d) Technical issues to be developed; e) Compatibility with hydrogen production system, etc. And for the second step, as the system optimization study among the selected system, sensitiveness to plant efficiency by changing the inlet and the outlet temperature of reactor core has been investigated from an economical and plant efficiency point of view.

The Component Test Facility: A National User Facility for Testing of High Temperature Gas-Cooled Reactor (HTGR) Components and Systems

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

Structural Integrity Code and Regulatory Issues in the Design of High Temperature Reactors

Since the 1980s, the ASME Code has made numerous improvements in elevated-temperature structural integrity technology. These advances have been incorporated into Section II, Section VIII, Code Cases, and particularly Subsection NH of Section III of the Code, “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400°F (760°C) to about 1742°F (950°C) where creep effects limit structural integrity, safe allowable operating conditions, and design life. Materials that are more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses. Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high temperature applications. Current ASME Section III design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking and corresponding criteria are needed for high temperature reactors. Subsection NH of Section III was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC, DOE and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. The design lifetime of Gen IV Reactors is expected to be 60 years. Additional materials including Alloy 617 and Hastelloy X need to be fully characterized. Environmental degradation effects, especially impure helium and those noted herein, need to be adequately considered. Since cyclic finite element creep analyses will be used to quantify creep rupture, creep fatigue, creep ratcheting and strain accumulations, creep behavior models and constitutive relations are needed for cyclic creep loading. Such strain- and time-hardening models must account for the interaction between the time-independent and time-dependent material response. This paper describes the evolving structural integrity evaluation approach for high temperature reactors. Evaluation methods are discussed, including simplified analysis methods, detailed analyses of localized areas, and validation needs. Regulatory issues including weldment cracking, notch weakening, creep fatigue/creep rupture damage interactions, and materials property representations for cyclic creep behavior are also covered.

Examination of Dust in AVR Pipe Components

Pebble Bed HTGR’s like the AVR in Ju¨lich have the advantage of continuous fuelling. However the multiple passes of the fuel pebbles through the core have the disadvantage that the pebble’s movement through the fuelling system and the core produces graphite dust. This dust is transported from the core to other parts of the primary circuit and deposits on components. Although previous experiments performed during AVR operation have given some insight into the dust particle size and activity, there is little information on the behaviour of the dust that was deposited in the system. The decommissioning of the AVR has provided the opportunity to sample and characterise such dust from a number of components and gauge the adhesion strength. From the side of PBMR Pty Ltd this opportunity is considered important to enhance the knowledge about dust characteristics before the PBMR Demonstration Power Plant (DPP) is operational and able to produce specific plant information through sampling and analysis. AVR GmbH has provided a number of pipes and joints for investigation of loose and bound dust. Phase 1 of the analysis was used to determine the best techniques to be used on larger items. No measurable loose dust could be collected. Thereupon rings were cut from a T-section and subdivided into eight segments. The surface of the untreated segments were photographed and documented by optical microscopy, the dose rates were measured and gamma-spectrometry performed. Following this a mechanical or chemical decontamination was carried out to remove and isolate the bound dust. The average isolated dust amount was about 2 mg/cm2. Both decontamination processes indicates a strong bonding of the dust surface layer. In the case of mechanical decontamination about 60% and by chemical decontamination about 95% of the radionuclide inventory could be removed. The contribution of removed metal needs to be investigated in more detail. The median number related particle size measured by optical microscopy was found to be in the range of 0.2 to 0.7 μm whereas the median weight related size is in the range of 0.8 to 1.5 μm. The initial results indicate that this dust sticks very strongly to the pipe surface. Phase 2 will concentrate on longer pieces of piping where hopefully more loose dust can be obtained and analysed. If the same strong bonding is observed the reason for this phenomenon needs to be explained and perhaps tested with non-active dust.

Providing Thermal-Hydraulic Boundary Conditions to the Reactor Code TINTE Through a Flownex-TINTE Coupling

TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time.

TH3D: A Three-Dimensional Thermal Hydraulic Tool for Design and Safety Analysis of HTRs

The institute of nuclear engineering and energy systems (IKE), University of Stuttgart, Germany has developed a new thermal hydraulic tool which can be used for three-dimensional thermal hydraulic analysis of pebble bed as well as block type HTRs. During nominal operation, the flow inside the gas-cooled High Temperature Reactor is essentially single-phase, compressible, and non-isothermal. So, at least one gas phase has to be considered beside the solid phase for thermal hydraulic analysis of HTRs. Each phase (e.g. solid, gas) is considered as a continuum which occupies only its respective fraction of the control volume. Thermal non-equilibrium is considered between phases and time dependent energy conservation equations for solid and gas phases are solved. Simplified momentum conservation equation for gas obtained from porous media approximation is solved along with the time dependent mass conservation equation. Provisions for simulating more than one gas component are available in this newly developed code TH3D which could be required for simulating some accident situations (e.g air / water ingress by pipes break). The interaction between phases is made by a set of constitutive equations which rely on semi-empirical correlations obtained from different experiments. Finite volume method with a staggered grid approach is used for spatial discretization and a fully implicit, time adaptive, multi step method is used for time-dependent discretization. A benchmark calculation which is oriented to the pebble type fuel reactor PBMR-400 and a 3D calculation were presented in HTR-2006 conference and will also be published in Nuclear Engineering and Design (NED) journal. In order to demonstrate the capabilities of TH3D for simulating all block type HTRs, a benchmark calculation which is proposed by IAEA CRP-3 and oriented to the Gas Turbine Modular Helium Reactor (GT-MHR) is performed. Calculations are performed for the steady state case (nominal operation) as well as for Loss of Forced Cooling (LOFC) with and without depressurization. The results obtained from TH3D are compared with the results obtained from several countries participated in this benchmark calculation program by using different code system. In this paper, results of this benchmark calculation and comparisons will be presented. A fuel model for pebble type fuel is implemented in TH3D where heterogeneity of heat production inside the fuel pebble is taken into account. The assumption of homogeneous heat production could be justified for steady state calculation or for slow transient but for fast transient calculation, the assumptions of homogeneous and heterogeneous heat production produce a huge difference for coupled thermal hydraulics and neutronics calculation. In order to show the capabilities of this newly developed code TH3D to couple with a neutronics system, it was coupled with a point kinetics model for a fast reactivity insertion case. In this case all control rods were withdrawn very quickly (with a velocity of 1 m/sec) to the end position. It was assumed that the scram signals were not activated when power or temperature was increased beyond a limiting value during this withdrawal process but the control rods system continued to be withdrawn up to the top position instead of getting down and the coolant flow was reduced by controlling the blowers. The neutronics feedback during this fast reactivity insertion case with homogeneous and heterogeneous fuel model will also be presented.

An Infiltration Manufacturing Process for Nuclear Fuels

The Infiltrated Kernel Nuclear Fuel (IKNF) process deposits nuclear fuel into the naturally occurring porosity in graphite. IKNF consists of infiltrating uranyl nitrate dissolved in an organic solvent into the graphite and then heat-treating the sample at low (<300°C) temperatures to remove the solvent and convert the uranyl nitrate to UO2. Complete conversion to UC2 can then be accomplished by heating to temperatures higher than 3000°C. IKNF is extremely flexible: it is appropriate for very high temperature applications and heating the infiltrated product to intermediate temperatures (higher than 900°C) produces nuclear fuel with a range of chemistries in the U-C-O system (similar to the current US TRISO fuel). It is probable that the process can also be used to produce fuel containing transuranics. It is believed that IKNF will be less expensive, more robust and more suitable for on-line quality monitoring than current fuel fabrication method. Graphite infiltration involves a few, easily measurable and controllable variables. It is reproducible and predictable.

Performance Comparison of VHTR Plants With Direct and Indirect Energy Conversion Cycles

This paper compared the performance of very high temperature reactor (VHTR) plants with direct and indirect closed Brayton Cycles (CBCs) and investigated the effect of the molecular weight of the CBC working fluid on the number of stages in and the size of the single shaft turbomachines. The CBC working fluids considered are helium (4 g/mole) and He-Xe and He-N2 binary mixtures (15 g/mole). Also investigated are the effects of using LPC and HPC with inter-cooling, cooling the reactor pressure vessel with He bled off at the exit of the compressor, and changing the reactor exit temperature from 700°C to 950°C on the plant thermal efficiency, CBC pressure ratio and the number of stages in and size of the turbo-machines. Analyses are performed for reactor thermal power of 600 MW, shaft rotation speed of 3000 rpm, and IHX temperature pinch of 50 °C.

Turbulent Scalar Flux Models for the Mixing of Hot Gases in the Swirling Flow of High-Temperature Reactor Lower Plenums

With recent progress in high-temperature pebble-bed reactor programs research focus has started to include more ancillary engineering issues. One very important aspect for the realisability is the mixing of hot and colder helium in the reactor lower plenum. Under nominal operating conditions, depending on core design, the temperature of hot gas leaving the core can locally differ up to 210° C. Due to material limitations, these temperature differences have to be reduced to at least ±15° C. Several reduced-size air experiments have been performed on this problem, but their applicability to modern commercially sized reactors is not certain. With the rise in computing power CFD simulations can be performed in addition, but advanced turbulence modeling is necessary due to the highly swirling and turbulent nature of this flow. The presented work uses the geometry of the German HTR-Modul which consists of an annular mixing channel and radially arranged ribs. Using the commercial CFD code ANSYS CFX, we have made detailed analyses of the complex 3D vortical flow phenomena within this geometry. Several momentum transport turbulence models, e.g. the classical k-e model, advanced two-equation models and Reynolds-Stress Models were compared with respect to their accuracy for this particular flow. In addition, the full set of turbulent scalar flux transport equations was implemented for modeling the three components of turbulent transport of enthalpy seperately and were compared with the standard turbulent Prandtl number approach. As expected from previous work in related fields of turbulence modeling, the differences in predicting the mixing performance between models were significant. Only the full Reynolds-Stress model coupled with the scalar flux equations was able to reproduce the experimentally observed reduction of mixing efficiency with increasing Reynolds number. The correct scaling of mixing efficiencies demonstrates that the utilized turbulence models are able to reproduce the physics of the underlying flow. Hence they could be employed for the scaling and optimization of the lower plenum geometry. The results also showed that the original geometry used for the HTR-Modul is insufficient to provide adequate mixing, and that hence a not sufficiently mixed coolant for future reactor designs might be an issue. Based on this work, an optimization for future lower plenum geometries has become feasible.