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.

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.

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.

Update on Overcoating and Compacting Activities for Coated Particles With 425 µm Kernels

The Advanced Gas Reactor (AGR) program is tasked with developing and qualifying fuel for the Next Generation Nuclear Plant (NGNP) [1, 2]. The first experiment, AGR-1, focused on TRISO coating 350 μm uranium oxide/uranium carbide (UCO) kernels and compacting them into a right circular cylinder fuel form using an overcoating and compacting process. The AGR-1 fuel compacts are currently being irradiated at the Advanced Test Reactor (ATR). The AGR-2 experiment will focus on overcoating and compacting TRISO coated 425 μm UCO kernels. This paper summaries the work that has been done to date on preparing to make AGR-2 compacts.

VHTR Core Preliminary Analysis Using NEPHTIS3 / CAST3M Coupled Modelling

Along with the GFR another gas-cooled reactor identified in the Gen IV technology roadmap, the VHTR is studied in France. Some models have been developed at CEA relying on existing computational tools essentially dedicated to the prismatic block type reactor. These models simulate normal operating conditions and accidental reactor transients by using neutronic [1], thermal-hydraulic, system analysis codes [2], and their coupling [3, 4]. In the framework of the European RAPHAEL project, this paper presents the results of the preliminary investigations carried out on the VHTR design. These studies aimed at understanding the physical aspects of the annular core and to identify the limits of a standard block type VHTR with regard to a degradation of its passive safety features. Analysis was performed considering various geometrical scales: fuel cell and fuel column located at the core hot spot, 2D and 3D core configurations including the coupling between neutronic and thermal-hydraulic. From the thermal analysis performed at the core hot spot, the capability to reduce the maximum fuel temperature by modifying the design parameters such as the fuel compact and the fuel block geometry was assessed. The best performances are obtained for an annular fuel compact geometry with coolant flowing inside and outside the fuel compact (ΔT > 50°C). The reliability of such design option should however be addressed with respect to its performance during the LOFC transient (the residual decay heat will be evacuated by radiation during the transient instead of conduction through graphite). As far as the fuel element geometry is concerned, a gain of approximately 50°C can be achieved by making limited changes on the fuel compact distribution in the prismatic block: reduction of the number of fuel compact in the outer ring of the fuel element where the average ratio between coolant channels and fuel compact is smaller. On the other hand, the adopted modifications should also be evaluated with respect to the maximum temperature gradient achieved in the fuel (amoeba effect). In the end, calculations performed on the full core configuration taking into account the thermal feedback showed that the radial positioning of the fuel elements allows to reduce significantly the power peaking factor and the maximum fuel temperature. The gain on the fuel temperature, which varies during the core irradiation, is in the range 100 – 150°C. Several modifications such as the increase of the bypass fraction and the replacement of a part of the graphite reflector by material with better thermal properties were also addressed in this paper.

VHTR-Based Systems for Autonomous Co-Generation Applications

As highly efficient advanced nuclear systems, Generation IV Very High Temperature Reactors (VHTR) can be considered in a variety of configurations for electricity generation and process heat applications. Simultaneous delivery of electricity, low-temperature process heat (for potable water production, district heating, etc.) and high temperature process heat (for hydrogen production, etc.) by a single cogeneration system offers unique deployment options as “all-in-one” power stations. This paper is focused on the VHTR-based systems for autonomous co-generation applications. The analysis is being performed within the scope of the U.S. DOE NERI project on utilization of higher actinides (TRUs and partitioned MAs) as a fuel component for extended-life VHTR configurations. It accounts for system performance characteristics including VHTR physics features, control options and energy conversion efficiencies. Utilization of TRUs in VHTRs is explored to stabilize in-core fuel compositions (core self-stabilization) leading to extended single-batch OTTO (Once-Through-Then-Out) modes of operation without intermediate refueling.

Advanced High Temperature Gas-Cooled Reactor Systems

Three systems have been proposed for advanced high temperature gas-cooled reactors (HTGRs): a supercritical carbon dioxide (S-CO2) gas turbine power conversion system; a new MicroChannel Heat Exchanger (MCHE); and a once-through-then-out (OTTO) refueling scheme with burnable poison (BP) loading. An S-CO2 gas turbine cycle attains higher cycle efficiency than a He gas turbine cycle due to reduced compression work around the critical point of CO2. Considering temperature lowering at the turbine inlet by 30°C through the intermediate heat exchange, the S-CO2 indirect cycle achieves efficiency of 53.8% at turbine inlet temperature of 820°C and turbine inlet pressure of 20 MPa. This cycle efficiency value is higher by 4.5% than that (49.3%) of a He direct cycle at turbine inlet temperature of 850°C and 7 MPa. A new MCHE has been proposed as intermediate heat exchangers between the primary cooling He loop and the secondary S-CO2 gas turbine power conversion system; and recuperators of the S-CO2 gas turbine power conversion system. This MCHE has discontinuous “S”-shape fins providing flow channels with near sine curves. Its pressure drop is one-sixth reference to the conventional MCHE with zigzag flow channel configuration while the same high heat transfer performance inherits. The pressure drop reduction is ascribed to suppression of recirculation flows and eddies that appears around bend corners of zigzag flow channels in the conventional MCHE. An optimal BP loading in an OTTO refueling scheme eliminates the drawback of its excessively high axial power peaking factor, reducing the power peaking factor from 4.44 to about 1.7; and inheriting advantages over the multi-pass scheme because of the lack of fuel handling and integrity checking systems; and reloading. Because of the power peaking factor reduction, the maximum fuel temperatures are lower than the maximum permissible values of 1250°C for normal operation and 1600°C during a depressurization accident.