Design Of High Temperature Gas-cooled Reactor Research Articles

Equilibrium core modeling of a pebble bed reactor similar to the Xe-100 with SCALE

As the nuclear industry moves towards licensing and constructing advanced reactors, new attention has been focused on the advanced reactor designs that have past operational experience, such as pebble-bed high-temperature gas-cooled reactors (PB-HTGRs). Pebble-bed reactor designs have many advantages, such as their higher operating temperatures and online refueling capabilities. However, high-fidelity computational modeling of pebble-bed reactor designs, from reactor startup to operation at equilibrium, is more challenging compared to conventionally fueled reactors due to the continuous movement of the fuel pebbles through the reactor during operation. In previous work at Oak Ridge National Laboratory (ORNL), the SCALE Leap-In method for Cores at Equilibrium (SLICE) was developed around tools within the SCALE code system. This iterative method can effectively generate pebble-bed reactor zone-wise fuel inventories at equilibrium core operation within a reasonable computational time. The objective of this work was to further verify the ORNL SLICE method and to investigate the impact of considering temperature profiles during the application of the method. The SLICE method was applied to a modular high-temperature gas-cooled reactor design based upon publicly available design specifications of the Xe-100 pebble-bed reactor. Upon comparing the results from the SLICE method to published literature, the differences in the eigenvalue keffective were on the order of several hundred pcm (percent millirho). To investigate one possible cause of these differences, a study looking at the sensitivity of the full-core equilibrium keffective and discharge nuclide inventory to temperature was performed by developing equilibrium cores of two additional temperature profiles. From this temperature study, differences on the order of hundreds of pcm for the full-core equilibrium keffective, and up to 15% difference for the discharge inventories were found. These results indicated the strong dependence on temperature that needs to be considered for future work in equilibrium modeling of PB-HTGRs.

Assessment of Reactor Physics Characteristics of Prismatic Fuel Concepts With a Hydrogen-Based Moderator for Use in a Fluoride Salt-Cooled Small Modular Reactor

Abstract Fluoride-salt-cooled high temperature reactor (FHR) effectively combines the solid fuel and moderator design of high-temperature gas-cooled reactor (HTGR) technology with the fluoride salt coolant (LiF-BeF2, FLiBe) of molten salt reactor (MSR) technology, enabling low-pressure (∼1 atm, 101.325 kPa), and high-temperature (∼700 °C) operations. The design and operational features of the FHR make it a potentially attractive option for a small modular reactor (SMR), provided that it can be modified and made physically small and operate at a low-enough power level (<350 MWth/<150 MWel). Most FHR-SMR designs use high-assay low enriched uranium (HALEU) fuel in the form of tri-structural isotropic (TRISO) fuel particles, combined with the use of a graphite moderator. However, there are alternative design concepts for an FHR-SMR that may offer superior performance characteristics, while utilizing an alternative fissile fuel supply option. In this exploratory study, lattice physics calculations were performed with Serpent to evaluate an alternative FHR-SMR prismatic fuel block design concept using coated annular fuel pellets instead of TRISO-particle fuel compacts, along with the use of hydrogen-based solid moderator rods made of 7LiH. In initial studies, it was found that fuel blocks with 120 moderator rods made of 7LiH tended to have large positive temperature reactivity coefficients (TRCs), which is undesirable for safety reasons. However, reducing the number of moderator rods to 90 or 54, while increasing the number of fuel rods and coolant holes led to low or negative temperature coefficients. For a prismatic fuel block design with 54-7LiH moderator rods, the isothermal temperature coefficient of reactivity (Isothermal TRC), with simultaneous changes in the fuel (F), graphite (G), hydrogen (H), and coolant (C) temperatures, ranges between −0.159 mk/K and −0.054 mk/K, depending on the operating temperature and fuel burnup. Such alternative FHR-SMR fuels could achieve a single-batch core life of ∼10 years with low enriched uranium (LEU) fuel, and ∼45 years with HALEU, in a 350-MWth reactor core.

TRISO burnup-dependent failure analysis in FHRs using BISON

In this study, we perform a study on TRIstructural ISOtropic (TRISO) Silicon Carbide (SiC) layer failure using the BISON fuel performance code. SiC thermomechanical behavior is analyzed during both steady-state and accident scenarios pertaining to a Fluoride-Salt-Cooled High-Temperature Reactor (FHR). A monodimensional single particle model is adapted from a previous analysis and run at 140 mW/particle for 500 days. The maximum burnup reached is 18.49% FIMA. A total of three accident scenarios are simulated at both fresh fuel conditions and following multiple irradiations at 2% FIMA increments up to 18.49% FIMA. The instances include an overcooling transient, a Control Rod Withdrawal (CRW) and a Loss of Forced Cooling Accident (LOFA). SiC hoop stress and failure probability predictions are evaluated for all simulated cases. A consistent tangential compressive stress state is found as a function of burnup for SiC. The maximum compressive hoop stress reaches approximately -400 MPa. The maximum SiC failure probability as a function of burnup is 2⋅10−13% at steady-state and no increases are predicted during transient simulations. It is concluded that the selected accident scenarios, under our modeling assumptions, are not predicted to pose a challenge to SiC integrity due to their relatively slow progression and low temperature increments, for both fresh fuel and at burnup values up to 18.49% FIMA. A 2D model is used, along with XFEM, to simulate a crack in the Inner Pyrolytic Carbon (IPyC). Conservative IPyC failure predictions are obtained with Weibull statistics approximately 38 days after beginning of irradiation. After crack formation, hoop stress concentrations are predicted in SiC using conservative assumptions, with failure probability increasing to 3.3⋅10−4%. 2D BISON simulations of the base irradiation are then carried out to the maximum burnup and transient conditions are subsequently applied. SiC maximum tensile hoop stress is predicted at the time of crack initiation, after which crack expansion reduces the predicted stress magnitude in the material. The results presented in the work are interpreted under the selected modeling assumptions and conclusions are based on a comparison with a parallel analysis on a High-Temperature Gas-Cooled Reactor design to evaluate the effects of power density and burnup on SiC performance.

Developing PCC and DCC integral effects test experiments at the High Temperature Test Facility

Among the Next-Generation Nuclear Plant (NGNP) designs, the High-Temperature Gas-Cooled Reactors (HTGRs) are very attractive, due to their inherent safety features, high power conversion efficiency, and potential of providing high-temperature process heat. To perform a thorough safety study and to license these types of reactors, sufficient information needs to be provided about the phenomena that occur during accident scenarios. While several experimental research efforts have been dedicated in the past to investigate accident scenarios, knowledge gaps still exist in the phenomena characteristic of pressurized and depressurized conduction cooldown (PCC/DCC) transients as well as for normal operation scenarios. This paper summarizes the Oregon State University High Temperature Test Facility (HTTF) test matrix, experimental campaign, and selected tests results. High Temperature Test Facility is a scaled Integral Test Facility (IET) that is capable of mimicking scaled dimensions and operational conditions of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The goal of the High Temperature Test Facility is to provide experimental data on the DCC, PCC and normal operating scenarios of the reference Modular High-Temperature Gas Cooled Reactor design. The DCC, PCC, mixing, heat up and cooldown tests described in this paper were performed at prototypical Modular High-Temperature Gas Cooled Reactor temperatures, scaled initial pressure conditions (∼200 kPa), and thermal power input of less than 70 kW. Presented test data show temperature distributions in the High Temperature Test Facility core, upper plenum, cross duct, or lower plenum. Based on these temperature profiles attempts to investigate stratified flow, natural convection flow, heat up, cooldown and mixing phenomena are made. Furthermore, this paper evaluates the performed test campaign in the light of the Very High Temperature Gas-cooled Reactor Phenomena Identification and Ranking Table (PIRT) and proposes experiments to complement the existing PCC/DCC testing database for the validation of the thermal-hydraulic codes.

Sub-cell radiation-conduction (SCRC) model in densely packed beds of mono-sized pebbles for HTGR under vacuum condition

The equivalent sub-cells of radiation and conduction are modeled for close packing structures of mono-sized pebbles. The aim is to calculate and analyze the average radiative and conductive heat transfer of high-temperature gas-cooled reactors (HTGRs) under different ranges of temperature to further assist in the design and operation of HTGRs. The sub-cell radiation-conduction (SCRC) model is established analytically according to the analysis of thermal resistance networks for radiation and conduction. It can be utilized to calculate the effective thermal conductivity (ETC) due to radiation and conduction in an HTGR under vacuum conditions. Comparisons of data of ETC obtained from the SCRC model, empirical models, and experiments are carried out under vacuum conditions. Comparisons of the non-dimensional radiation exchange factor (REF) between the SCRC model and other occasionally used models are also given. The comparisons show that the SCRC model is generally in accord with the commonly used models and the experimental results. The results demonstrated that the established SCRC model is expected to be able to describe the radiation and conduction in closely packed pebble beds under wide temperature ranges and vacuum conditions reasonably.

ASSESSMENT OF RADIOLOGICAL IMPACTS FROM POSTULATED ACCIDENT CONDITIONS OF HTGR: A CASE STUDY IN SERPONG NUCLEAR AREA

High Temperature Gas-cooled Reactor (HTGR) design has an improved safety which depends on its TRISO coated fuel particles that are considered not to be damaged even in accident condition. However, the radiological impacts from accident condition in HTGR is still important to be assessed. This research is aimed to perform radiological impacts assessment of two postulated accidents of HTGR, which are depressurization and water ingress accident. As a case study, a 10-MWTh pebble-bed HTGR design named Reaktor Daya Eksperimental with the planned site located in Serpong Nuclear Area was chosen. The source terms from the accident conditions were estimated using mechanistic source term model and the dose consequences were calculated using PC-COSYMA. The input data for PC COSYMA, which are meteorological, population distribution, agricultural and local farm data, were compiled based on the site data of Serpong Nuclear Area. The radiological impacts were assessed based on individual and collective doses. The results showed that the highest dose will be received by the community within a radius of 250 m to the south from the reactor. It was also found that these accidents only cause minor radiological impacts which do not meet the criteria for any countermeasures (iodine thyroid blocking, sheltering, evacuation, food ban, decontamination, and relocation).

Neutron importance based lattice-to-core projection technique and its application for HTGR design using Monte Carlo method

In general, a nuclear core design process is divided into two subsequent phases, namely the fuel lattice design phase, and followed by the full core design phase such as for the light water reactor (LWR). The optimal design parameters obtained from the fuel design phase can be used in a straightforward manner in the full core design phase since the neutron mean free path is much smaller than the fuel assembly dimensions. Unfortunately, the above-mentioned favorable situation may not be applicable for a high-temperature gas-cooled reactor (HTGR) design process. In this paper, for the HTGR design process, we propose a new technique based on the neutron importance concept to make a projection of the optimal design parameters obtained from the fuel design phase to the full core design phase so that the optimization works required in the full core design phase can be significantly reduced. The proposed technique is implemented in the Japanese High-Temperature Engineering Test Reactor to find the optimal fuel composition that can achieve an average core discharge burnup of 80 GWd/tU using single-batch and multi-batch refueling schemes. Important core neutronics parameters, such as reactivity coefficient and power profiles, are also discussed. Moreover, the impact of the axial shuffling scheme on the fuel compact fuel temperature is also analyzed.

Simulation of the HTR-10 Operation History With the PANGU Code

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-cooled Reactor (HTGR) in China, which was operated from January 2003 to May 2007. The HTR-10 operation history provides very important data for the validation of HTGR codes. In this paper, the HTR-10 operation history is simulated with the PANGU code, which has been recently developed for HTGR reactor physics analysis and design. Models and parameters are constructed based on the measured data of the actual conditions. The simulation results agree well with the measurements in all steady-state power periods. The discrepancy of keff is generally below 0.5%, and the discrepancy of coolant outlet temperature is generally below 5°C. It is also figured out that the burnup of graphite impurities has considerable influence on the keff at the end of the operation history, which can cause over 1.5% discrepancy when neglecting the burnup of graphite impurities. By this work, the PANGU code’s applicability in actual HTGR fuel cycle simulations is demonstrated.

Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

Three-dimensional modeling and loss-of-coolant accident analysis of high temperature gas cooled reactor

The high temperature gas-cooled reactor (HTR) is well known due to its prominent inherent safety, high efficiency and low carbon environmental protection. Accurate and reliable thermal hydraulic analysis is of great significance to the system design of HTR. Based on the coolant flow and heat transfer process in the core, this study establishes a three-dimensional coupled model for the core of HTR, and the temperature variation in the core is simulated. The model is verified by comparing the calculated results with those published data by THERMIX software. After that, the models are employed to study thermal hydraulic characteristics under the loss of coolant accident. The results show that the maximum relative error of the calculated results with published data are 2.1% under full power, the three-dimensional model can be employed to simulate the flow and heat transfer process in the primary circuit of HTR. The calculation of 50% of full power shows that the model fully considers the influence of convective heat transfer and radiative heat transfer between the core of the pebble bed and the core side wall and the cold air chamber, making numerical calculations closer to actual condition. It is found when approximately 20% of the coolant is lost, the outlet temperature of helium reaches 1140 K, which exceeds the helium outlet temperature limit (1126 K). In addition, the inherent safety of an HTR be verified under transient conditions of natural convection.

Physics study of block/prismatic-type HTGR design option for the Indonesian Experimental Power Reactor (RDE)

In our previous works, we have investigated the pebble-bed type high-temperature gas-cooled reactor (HTGR) as a candidate design for the 10 MWth Indonesian Experimental Power Reactor (“Reaktor Daya Experimental,” RDE). The once-through-then-out (OTTO) and multipass fuelling schemes with both uranium and thorium fuel cycles were considered. The heavy metal loading per ball and fissile enrichment as free design parameters were surveyed to obtain the optimum fuel compositions (the smallest fissile requirement per unit energy generated) for an average burnup target of 80 MWd/tU. Steady-state thermal-hydraulic analyses were also conducted, and no critical challenge was found since the reactor power density is low. Geographically the RDE site and the country, in general, are located in the so-called Ring of Fire (Circum-Pacific Belt) which is vulnerable against frequent seismic events. The nature of a block- or prismatic-type HTGR does not allow large and random movement of its fuel elements during a large seismic event which makes this type of HTGR relatively safer for these regions. The design also allows control and safety rods (absorbers) to be located inside the core by providing holes in graphite blocks which ensure the sub-criticality during an abnormal event. In the present work, under the same RDE design requirements and constraints, the block/prismatic-type 10 MWth HTGR is studied based on the Japanese 30 MWth High-Temperature Engineering Test Reactor (HTTR). Although the design study emphasized on the optimal fuel composition in the reactor physics lattice phase, a preliminary result of the full core reactor physics analytical work is given.

The Evaluation of the High Temperature Gas Cooled Reactor Safety to Fulfill the Requirement of the Next Generation Nuclear

High temperature gas cooled reactor (HTGR) has been considered to be the most promising option to meet energy demands in the future. It has also been selected as the next generation nuclear plant. The primary safety requirement of the next generation nuclear plant design is to limit radioactive material releases to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary. The purpose of this study is to evaluate the safety design of HTGRs in order to fulfill the requirement of the next generation nuclear plant. To achieve this objective, inherent safety features, fundamental safety functions, and confinement functions realized into the design of HTGRs are comprehensively evaluated. It is found that design provisions of HTGRs can fulfill the intention of keeping radionuclides at their original sources. The layers of the coated fuel particles are very robust to retain nuclear fission products for all foreseeable reactivity events. There will be no possibility of radioactive materials to be released even though related safety systems and operator intervention are not involved in the recovery actions. This design has complied with the requirement of the next generation nuclear plant, which is to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary.Keywords: High temperature gas cooled reactor, inherent safety features, fundamental safety functions, confinement functions, next generation nuclear plant

Self-shielding effect of double heterogeneity for plutonium burner HTGR design

The investigation on self-shielding effect of double heterogeneity for plutonium burner High Temperature Gas-cooled Reactor (HTGR) design has been performed. Plutonium burner HTGR designed in the previous study by using the advantage of double heterogeneity to control excess reactivity. In the present study, the mechanism of the self-shielding effect is elucidated by the analysis of burn-up calculation and reactivity decomposition based on exact perturbation theory. As a result, it is revealed that the characteristics of burn-up reactivity are determined by resonance cross section peak at 1 eV of Pu-240 due to the surface term of background cross section, this is, the characteristics of neutron leakage from fuel lump and collision to a moderator. Moreover, significant spectrum shift is caused during the burn-up period, and it enhances reactivity worth of Pu-239 and Pu-240 in EOL.

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.

Assessment of RELAP5 code model to simulate U-shaped heat pipe performance for heat sink

One of important component in a high temperature gas cooled reactor (HTGR) is reactor cavity cooling system (RCCS). The function of the RCCS is to remove the heat going out from the reactor pressure vessel. On the other hand, the use of passive system is highly considered in the nuclear reactor design, including in the HTGR design. The heat pipe is one of the passive apparatus. This current study deals with the research on the use of heat pipe to the RCCS design. As part of the study, the performance of U-shaped heat pipe will be assessed both through an experiment and numerical simulation methodology. The objective of this study is to assess the validity of the RELAP5 code model to simulate a U-shaped heat pipe performance. A small U-shaped heat pipe with fins in the condenser part is modelled using the RELAP5/SCDAP code. The result of simulation shows that the heat pipe could removed the heat received in the evaporator to the environment through the condenser of that heat pipe. The phenomena of evaporation and condensation are identified during heat removal. However, the steady state condition is not obtained until the end of the calculation. The calculated thermal resistance of the heat pipe is approx. 0.01258 °C/W. It is concluded that the model could simulate reasonably the heat pipe performance. The model needs to be improved, especially for the finned tube condenser.

Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

Fully ceramic microencapsulated (FCM) fuel may enhance the already strong inherent reactor safety characteristics of a conventional High-Temperature Gas-Cooled Reactor (HTGR). FCM fuel uses an SiC matrix that exhibits higher stability under irradiation with limited swelling relative to a conventional graphite matrix. Additionally, the SiC matrix is expected to exhibit improved mechanical performance relative to graphite. The feasibility of FCM-fueled HTGRs has been explored in a previous study (Lu et al., 2018), and three FCM fuel concepts were identified to be able to maintain the cycle length of the reference conventional General Atomics 350 MWt prismatic modular HTGR.In this work, the previous investigation of the reactor core safety characteristics of HTGRs fueled by these three FCM fuels was significantly extended to detailed thermal hydraulics and neutronics analysis of normal operation and design basis accident scenarios. We found that these FCM-fueled cores would have a maximum fuel temperature ∼20–40 K higher than that of the reference HTGR core at the beginning of cycle (BOC) and ∼30–60 K higher at the end of cycle (EOC) during normal operating conditions. Pressurized (P-) and depressurized (D-) loss of forced cooling (LOFC) accidents as well as the control rod withdrawal accident were explored. The FCM-fueled cores would at most have a maximum fuel temperature ∼30–50 K higher than that of the reference core in a P-LOFC accident, ∼10–30 K in a D-LOFC accident and ∼30–50 K in a rod withdrawal accident. The conclusion of the study is that FCM fuel has a small impact on the reactor performance and safety characteristics during normal operation and design basis accident conditions within the existing and generic 350 MWt prismatic modular HTGR design we considered.Additionally, a fuel cycle cost comparison was also performed in the current study. We found that the FCM fuels may be up to 40%–74% more expensive than the conventional fuel. Further analysis of the economics of FCM fuel are needed, because the uncertainties in the assumptions for this study are high.

Burnup characteristics analyses of graphite impurities in HTGR fuel element

In the fuel element of high-temperature gas cooled reactors (HTGRs), there exist various impurities in the graphite matrix, which are mainly originated from the raw materials, such as naturally occurring graphite, artificial graphite and resin. These impurities can reduce the excess reactivity of the reactor core, both in the initial criticality and during the overall operation history. In the practical physics design of HTGR, the equivalent boron content (EBC) method is usually used to describe the graphite impurities, where the impurity isotopes are simply converted to some amounts of boron by preserving the neutron absorption. Although it is found to be effective in criticality calculations, EBC may introduce errors in burnup analyses. This paper investigates the burnup characteristics of the graphite impurities. According to their behaviors during burnup, the impurities are classified into three categories, i.e. the high-burnable isotopes, low-burnable isotopes and unburnable isotopes. Moreover, a revised equivalent boron content (REBC) is proposed to describe the absorption rates of impurities, which can improve the computational accuracy in the burnup analyses of HTGR fuel.

Nuclear and thermal feasibility of lithium-loaded high temperature gas-cooled reactor for tritium production for fusion reactors

A high-temperature, gas-cooled reactor (HTGR) is proposed as a tritium production device that has the potential to produce a large amount of tritium using the 6Li(n,α)T reaction without major changes to the original reactor core design. In an HTGR design, generally, boron is loaded into the core as a burnable poison to suppress excess reactivity. In this study, lithium is loaded into the HTGR core aiming to produce thermal energy and tritium simultaneously and is loaded instead of boron as a burnable poison. The nuclear characteristics and fuel temperature were analyzed to confirm the nuclear and thermal feasibility of a lithium-loaded HTGR. It was shown that the analysis results satisfied the design requirements and hence the nuclear and thermal feasibility was confirmed for a lithium-loaded HTGR that produces thermal energy and tritium.

Fully ceramic microencapsulated fuel in prismatic high temperature gas-cooled reactors: Analysis of reactor performance and safety characteristics

Advanced nuclear reactor technologies have the potential to expand the missions of nuclear energy while reducing carbon emissions. This paper presents scoping reactor physics and thermal hydraulics analysis of a high temperature gas-cooled reactor (HTGR) using the fully ceramic microencapsulated (FCM) fuel form, and demonstrates the feasibility of FCM fueled HTGRs. FCM fuel consists of tristructural isotropic (TRISO) coated fuel particles embedded in a matrix of silicon carbide (SiC). The potential advantages of FCM fuel, which uses a monolithic SiC matrix, over conventional HTGR fuels with a carbon-based matrix include: a long refueling interval; high stability of the SiC matrix under irradiation with limited swelling; high fission product retention of the fuel form, with the SiC matrix acting as an additional barrier to fission product release; and enhanced oxidation resistance during normal operation and air ingress accidents. In addition, the literature shows that the effective thermal conductivity of SiC fuel compacts and conventional HTGR compacts are expected to be similar.The key finding of this study is that FCM fuel, within the form factor of a typical General Atomics prismatic graphite block, exhibits similar fuel cycle performance to conventional HTGR fuel. The reactor cycle length, discharge burnup, and natural resource utilization are similar. However, the reduced moderation in the FCM designs considered here does marginally reduce the discharge burnup, and therefore natural resource utilization, versus the reference HTGR design.The hardened neutron flux spectrum resulting from the SiC matrix, which displaces carbon from the core, requires a slightly higher packing fraction of conventional uranium oxy-carbide (UCO) fuel kernels or the use of higher density uranium mononitride (UN)-based fuel kernels. These options will marginally increase the decay power, because they harden the neutron flux energy spectrum and increase the density of 238U in the fuel. In one case considered, this will increase the absorption of neutrons in 238U, and the resultant impact of 239Np isotope on the decay power. The Doppler coefficients normalized per total fuel heat capacity are weaker in the FCM-fueled designs than in the reference HTGR design. This impacts the energy deposition in a control rod ejection accident, and hence the design of potential transient tests of these fuel forms. In addition, analyses of loss-of-forced cooling accidents indicate that the fuel temperature during these design basis accidents are up to ∼30 °C higher with FCM fuel than with conventional HTGR fuels due to the increased decay power.

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.