High Temperature Gas-cooled Reactor Concept Research Articles

Preliminary study of a small high-temperature gas-cooled reactor (HTGR) concept with MgO–BeO moderators

Graphite plays a crucial role as a moderator and reflector in high-temperature gas-cooled reactors (HTGRs). However, under high-irradiation conditions, graphite exhibits microcracking within the operational period. Studies have been investigated the potential of composite-based materials for replacing graphite in HTGRs. This study focused on a magnesium oxide (MgO)-based composite material as the host matrix and a homogeneously distributed beryllium oxide (BeO) as the entrained moderating phase and investigated the feasibility of the MgO–BeO as the new moderator in a small HTGR to achieve high burnup performance. In this study, the conceptual design of the HTR50S was selected as one of the candidates for the small HTGR concept. Burnup calculations and safety evaluation in HTR50S design were performed. The Monte Carlo MVP 3.0 and MVP-BURN codes were used in this study for neutronic calculations. Results revealed that a high burnup of 80 GWd/t can be achieved using a fuel composition of 6 kg heavy metal per fuel block with 17 wt% of 235U enrichment. Furthermore, a negative temperature coefficient of reactivity was achieved during the operation period.

FUEL CYCLE IMPLICATIONS OF DEPLOYING HTGRS IN HYBRID ENERGY SYSTEMS AS RESERVE POWER GENERATION IN ONTARIO

Nuclear power plants could potentially be deployed in a type of nuclear hybrid energy system (NHES) in which their power is used primarily to drive an industrial process but can be diverted to meet demands for electricity when needed. The purpose of this study is to analyze the effects of deploying NHESs as reserve power for the transmission grid in Ontario on the overall Canadian fuel cycle. In this scenario, the fuel cycle demands of 2 high-temperature gas-cooled reactor (HTGR) concepts are analyzed with respect to costs, resource consumption, and enrichment requirements. One HTGR concept is a 30 MW-thermal (MWth) reactor that is based on the UBattery concept, and the other is the Xe-100, which is a 200 MWth reactor. Calculations indicate that such a deployment of HTGRs would have a substantial effect on the fuel cycle in Canada. In particular, NU and enrichment demands would be greatly affected. Beginning this HTGR deployment in the year 2030 would more than double the annual NU demands in Canada, and deplete the uranium resources with extraction costs of <$80/kgU by the year 2142. The uranium enrichment demands of this fleet would be >35% of the US capacity for uranium enrichment.

NEUTRONIC BENCHMARK ON HOLOS-QUAD MICRO-REACTOR CONCEPT

The Holos-Quad micro-reactor concept is proposed by HolosGen LLC for civilian applications to generate 22 MWt with a lifetime of approximately 8 effective full power years (EFPYs). The design is based on a very innovative high-temperature gas-cooled reactor concept using four Subcritical Power Modules (SPMs) that fit into one commercial 40-foot transport ISO container. Neutronics benchmarks were developed based on a preliminary version of the Holos-Quad design to confirm feasibility of the neutronics design of this non-traditional high temperature gas-cooled (HTGR)-type micro-reactor concept. Calculations were performed using Monte Carlo codes (SERPENT and OpenMC) as well as the PROTEUS high-fidelity deterministic code for two exercises based on a unit cell model and a full core model. The results obtained showed good agreement with all of the evaluated parameters for the unit cell problem and the full core problem. SERPENT and OpenMC display consistently good agreement in eigenvalue within 150 pcm for the unit cell benchmark and less than 270 pcm for the full core benchmark. PROTEUS eigenvalues showed relatively larger differences but still reasonable agreement with the SERPENT solutions. For the depletion benchmark, the observed eigenvalue differences between SERPENT and OpenMC were within 300 pcm throughout the depletion up to 100 GWd/MT, ensuring that the two codes used equivalent input parameters such as the recoverable heat values for fission reactions. This benchmark exercise confirms the neutronic feasibility and core performance of the preliminary Holos-Quad design.

Influences of the ZrC coating process and heat treatment on ZrC-coated kernels used as fuel in Pu-burner high temperature gas-cooled reactor in Japan

ABSTRACT The concept of Pu-burner high temperature gas-cooled reactor (HTGR) has been proposed to more safely reduce the amount of recovered Pu. In the Pu-burner HTGR concept, coated fuel particles with ZrC-coated yttria-stabilized zirconia containing PuO2 (PuO2-YSZ) kernels are employed for very high burn-up and high nuclear proliferation resistance. The role of ZrC layer is that of oxygen getter. CeO2-YSZ kernels were fabricated to simulate the PuO2-YSZ kernel and coated with a ZrC layer. In this study, we clarified that both Ce-rich grains and Zr-rich grains were densely distributed in surface regions of the as-fabricated CeO2-YSZ kernel. However, we have already clarified that the surface region of the CeO2-YSZ kernel coated with a ZrC layer was porous and mainly consisted of Zr-rich grains. These experimental results confirmed that Ce-rich grains were selectively corroded during the ZrC coating process. Then, the chemical stability of Zr-rich grains would be higher than that of Pu-rich grains. Thus, it would be more difficult to extract Pu from PuO2-YSZ kernels (in which almost all grains are Zr-rich) than from PuO2-YSZ kernels (in which many Pu-rich grains are included). Influences of the sintering of fuel compact on the microstructure of the ZrC-coated CeO2-YSZ kernel is also reported.

Microstructures of ZrC coated kernels for fuel of Pu-burner high temperature gas-cooled reactor in Japan

Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing the amount of recovered Pu. In the Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO2 (PuO2-YSZ) kernel and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. Japan Atomic Energy Agency (JAEA) has carried out ZrC coatings of kernels simulating PuO2-YSZ kernels. Kernels which simulated PuO2-YSZ kernels were YSZ containing CeO2 (CeO2-YSZ) kernels fabricated by Nuclear Fuel Industry, Ltd. (NFI) or commercially available YSZ kernels. Ce was used as simulating element of Pu. In this manuscript, microstructures of ZrC coated CeO2-YSZ or YSZ kernels are reported, and future direction of development of Pu-burner HTGR is discussed. CeO2-YSZ kernel consisted of Ce-rich grains and Zr-rich grains. Observation results indicated CeO2-YSZ kernel was corroded during ZrC deposition process. Then Ce can be extracted from CeO2-YSZ kernels. If Pu can also be extracted from PuO2-YSZ kernels, that PuO2-YSZ kernel is undesirable in view of nuclear proliferation resistance. Fuel design and/or reactor core design of Pu-burner HTGR must be revised, if it is very hard to fabricate PuO2-YSZ kernels, with composition used in the present design of Pu-burner HTGR, and from which Pu is hard to be extracted. Observation results also indicated the possibility that only Ce-rich grains were corroded. If Pu can be extracted only from Pu-rich grains, PuO2-YSZ kernels containing no or only a small amount of Pu-rich grains, should be fabricated. ZrC layer was bound to YSZ kernel, but was completely debonded after heat treatment simulating sintering of fuel compacts. Remarkable grain growth occurred both in ZrC layer and YSZ kernel, after heat treatment. Voids were distributed in ZrC layer both in grain and on grain boundary.

Creep Effects on Design Below the Temperature Limits of ASME Section III Subsection NB

Some recent studies of material response have identified an issue that crosses over and blurs the boundary between ASME Boiler and Pressure Vessel Code Section III Subsection NB and Subsection NH. For very long design lives, the effects of creep show up at lower and lower temperature as the design life increases. Although true for the temperature at which the allowable stress is governed by creep properties, the effect is more apparent, e.g., creep effects show up sooner, at local structural discontinuities and peak thermal stress locations. This is because creep is a function of time, temperature, and stress, and the higher the localized stress, the lower in temperature creep begins to cause damage. If the threshold is below the Subsection NB to NH temperature boundary, 700°F for ferritic steels and 800°F for austenitic materials, then this potential failure mode will not be considered. Unfortunately, there is no experience base with very long lives at temperatures close to but under the Subsection NB to NH boundary to draw on. This issue is of particular interest in the application of Subsection NB rules of construction to some high temperature gas-cooled reactor concepts. The purpose of this paper is, thus, twofold: one part is about statistical treatment and extrapolation of sparse data for a specific material of interest, SA-533 Grade B Class 1; the other part is about how these results could impact current design procedures in Subsection NB.

Development of GAMMA Code as an Integrated Code Analyzing a Coupled Very High Temperature Gas-Cooled Reactor and Hydrogen Production Plant

The very high temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 1173 K (900°C) and operational fuel temperatures above 1523 K (1250°C). The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the high-temperature gas-cooled reactor concepts have sufficiently high temperatures to support process heat applications, such as hydrogen production, tar sands, oil shale, desalination, or cogenerative processes, the VHTR’s higher temperatures can be detrimental to safety if a loss-of-coolant accident occurs and causes the mechanical strength degradation of the supporting graphite in the lower plenum. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion or density-gradient–driven stratified flow phenomena and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Therefore, there was a need to develop a computer code that can be used for VHTR air ingress-related graphite oxidation analyses. Prior to the start of the Republic of Korea/United States International Nuclear Energy Research Initiative collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate an air ingress phenomenon in the VHTR. Therefore, we have worked for the past 3 yr on developing and validating advanced computational methods for simulating air ingress in the VHTR. The Idaho National Laboratory is developing a system integration model of VHTR and hydrogen production plant. GAMMA code is being considered to be an integrated computer tool to analyze the thermal hydraulics of the coupled plant. Computer models for a high-temperature steam electrolysis (HTSE) process were developed and were implemented in an overall system process optimization code, HYSYS. The HTSE model will be implemented into GAMMA code as the integrated computer tool.This paper describes the governing equations and numerical methods used in GAMMA code and presents a portion of verification of the GAMMA code along with turbomachinery models and HTSE models that will be linked to GAMMA code.

HTGR reactor physics and fuel cycle studies

The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the “HTR-N” project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides. These studies were mainly concerned with the investigation and intercomparison of the plutonium and actinide burning capabilities of a number of HTGR concepts and associated fuel cycles, with emphasis on the use of civil plutonium from spent LWR uranium fuel (first generation Pu) and from spent LWR MOX fuel (second generation Pu). Besides, the “HTR-N” project also included activities concerning the validation of computational tools and the qualification of models. Indeed, it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel. These validation and qualification activities have been centred round the two HTGR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3D diffusion theory codes yield acceptable results. Special attention, however, has to be given to the modelling of neutron streaming effects. For the HTR-10 the analyses focused on first criticality, temperature coefficients and control rod worth. Also in these studies a good correspondence between calculation and experiment is observed for the 3D diffusion theory codes.

Examining the Inherent Safety of Prism, Safr, and the Mhtgr

Three advanced design concepts, including two liquid-metal-cooled reactors (LMRs), the Power Reactor Inherently Safe Module (PRISM) and the Sodium Advanced Fast Reactor (SAFR), and a high-temperature gas-cooled reactor (HTGR) are discussed and compared. Each provides inherent or passive safety to improve system safety. The focus is on two primary objectives: reactor shutdown and shutdown heat removal. The LMR and HTGR concepts rely on inherent reactivity feedback to provide an inherent reactor response under a failure-to-scram condition; SAFR also provides a passive shutdown system using Curie point magnets (the self-actuated scram system). For shutdown heat removal, the LMR and HTGR designs rely on passive air cooling of the reactor vessel as the ultimate safety-grade system.