High-temperature Gas-cooled Reactor Safety Research Articles

Understanding Ag liquid migration in SiC through ex-situ and in-situ Ag-Pd/SiC interaction studies

The effective containment of fission products (FPs) within tri-structural isotropic (TRISO) fuel particles is crucial for the safety and efficiency of High Temperature Gas-cooled Reactors (HTGRs). Combining multi-scale (µm to nm) and multi-dimensional (2D and 3D) analysis, this study focuses on the migration mechanisms of silver (Ag), for which the Ag-110 m FP isotope is radiotoxic, facilitated by reactions with palladium (Pd) alloys and the silicon carbide (SiC) layer at elevated temperatures (1300–1500 °C). Three primary pathways for Ag migration are identified: (1) diffusion in solid palladium silicide to form (Pd,Ag)2Si, (2) infiltration through cracks and pores in the carbon phase, and (3) liquid phase migration along SiC grain boundaries. Especially at temperatures above the melting point of Pd2Si (1404 °C), a ‘dissolution-recrystallisation’ mechanism is proposed of the Ag-Pd-Si liquid phase migrating along SiC grain boundaries. The study employs state-of-the-art in-situ heating transmission electron microscopy (TEM) techniques to directly observe the dynamic migration processes of Pd silicide within SiC, providing unprecedented insights into the liquid behaviour in TRISO fuel. These findings highlight the significant role of liquid phases in determining the transport of FPs through the TRISO-SiC which is vital for developing strategies that enhance the safety and efficacy of HTGRs.

Management of Radioactive Waste from HTGR Reactors including Spent TRISO Fuel—State of the Art

In light of the increasing demand for energy sources in the world and the need to meet climate goals set by countries, there is growing global interest in high temperature gas cooled reactors (HTGRs), especially as they are known to be inherently safe nuclear reactors. The safety of HTGRs results, among other, from the nature of the nuclear fuel used in them in the form of coated TRISO particles (tri-structural-isotropic) and the reduction of the total amount of radioactive waste generated. This paper reviews numerous methods used to ensure the sustainable, feasible management and long-term storage of HTGR nuclear waste for the protection of the environment and society. The types of waste generated in the HTGR cycle are presented as well as the methods of their characterization, which are important for long-time storage and final disposal. Two leading nuclear fuel cycle strategies, the once-through cycle (direct disposal or open cycle) and the twice-through cycle (recycling or partially closed cycle), are discussed also in relation to TRISO spent fuel. A short review of the possibilities of treatment of TRISO spent nuclear fuel from HTGR reactors is made.

Improvement and validation of RELAP5/SCDAP code on evaluating the graphite behavior during oxidation process

RELAP5/SCDAP code has been improved by incorporating relevant chemical reaction models to evaluate the graphite oxidation behaviour in High Temperature Gas-cooled Reactor (HTGR). The developed code was validated against three individual experiments with different emphasis. It could be concluded that the developed Burn-off factor (M(B)) correlation showed a better performance on fitting the experimental data than the model proposed from the previous researches. Besides, the variation of the fractions of each gas species after the chemical reactions show a good agreement with the experimental results. Moreover, the temperature transient during the graphite oxidation was also reasonably reproduced with acceptable deviation. All in all, after the improvement, RELAP5/SCDAP code has a better capacity in evaluating the graphite behavior during the oxidation conditions, which improves its reliability on HTGR safety analysis.

DEM simulation for separating coated fuel particles by inclined vibrating plate

Odd-shaped coated nuclear fuel particles are the main reason for the failure in the manufacture of fuel element, and the sphericity separation of fuel particles is one of the most significant process to improve the safety of high-temperature gas-cooled reactor (HTGR). The separation of particles by inclined vibrating plate (IVP) is based on the different friction force acted by the distinct shapes of particles. In this paper, Discrete element method (DEM), which is a numerical simulation method for analyzing and solving the dynamic properties of discrete systems, is used to investigate the separation process on the IVP. In view of the realistic fuel particles are all non-spherical particles with irregular shape, the super-ellipsoid model is utilized to describe the particles in the DEM simulation. A method to evaluate the separation efficiency is proposed in this paper to quantify the numerical simulation results of particle sphericity separation. The relationship among the operating parameters such as amplitude, frequency and inclination angle is combed, and the selection of the optimal value of these parameters are obtained. The effects of feeding-rate and height of feeder are investigated, and the particle collision is analyzed to explain the effect of feeding-rate. A set of optimized parameters are proposed, and it is identified that the DEM simulation results received by using these parameters meet the requirement of the fabrication process of coated fuel particles.

A Comparative Study on Safety Design Requirements between HTGR and LWR

SSR–2/1 defines safety objectives, safety principles, and safety concepts, which shall be met by designs of nuclear power plants (NPPs). It provides NPP designs with an up to date level of safety to help organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning. General safety design requirements in the SSR–2/1 are defined based on the designs and technologies of light water reactors (LWRs). From the design and technology points of view, high temperature gas-cooled reactors (HTGRs) are very different from LWRs. In HTGR, the coated particle fuel is the strongest, ultimate and primary fission product barrier as opposed to the containment building in LWRs. Hence, maintaining the integrity of the coated particle fuel should be the focus of the HTGR safety design aspect. Consequently, the existing known and stable safety design requirements in SSR–2/1 are not well suited for HTGRs. The purpose of this paper is to comparatively study safety design requirements between HTGR and LWR. To achieve this objective, four important safety design requirements in the SSR–2/1, i.e. design extension condition, emergency core cooling system, functional containment, and accident management, are selected to be discussed. The study concluded that a tailored set of safety design requirements derived from those defined in SSR–2/1 should be developed by considering HTGR specific characteristics.

Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65g in AVR.

Investigation of Countermeasure Against Local Temperature Rise in Vessel Cooling System in Loss of Core Cooling Test Without Nuclear Heating

In the loss of core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of the reactor core is stopped without inserting control rods into the core and, furthermore, without cooling by the vessel cooling system (VCS) to verify safety evaluation codes to investigate the inherent safety of high-temperature gas-cooled reactor (HTGR) be secured by natural phenomena to make it possible to design a severe accident-free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel (RPV) by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water-cooling tube without thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from helium gas circulators (HGCs) by stopping water flow in the VCS, the local higher temperature position was specified in the uncovered water-cooling tube of the VCS, although the temperature was sufficiently lower than the maximum allowable working temperature, and the natural circulation of water had an insufficient cooling effect on the temperature of the water-cooling tube below 1°C. Then, a new safe and secured procedure for the loss of core cooling test was established, which will be carried out soon after the restart of HTTR.

Graphite dust resuspension in an HTR-10 steam generator

Graphite dust has an important effect on the safety of high-temperature gas-cooled reactors (HTR). The flow field in the steam generator was studied by the computational fluid dynamics (CFD) method, with the results indicating that the friction velocity in the windward and the leeward of the heat transfer tubes is relatively low and is higher at the sides. Further analysis of the resuspension of graphite dust indicates that the resuspension fraction reaches nearly zero for particles with a diameter less than 1μm, whereas it will increases as the helium velocity in the steam generator increases for particle size larger than 1μm. Moreover, the resuspension fraction increases as the particle size increases. The results also indicate that resuspension of the particles with sizes larger than 1μm exhibited obvious differences in different parts of the steam generator.

Automatic X-ray inspection for escaped coated particles in spherical fuel elements of high temperature gas-cooled reactor

As a core unit of HTGRs (high-temperature gas-cooled reactors), the quality of spherical fuel elements is directly related to the safety and reliability of HTGRs. In line with the design and performance requirements of the spherical fuel elements, no coated fuel particles are permitted to enter the fuel-free zone of a spherical fuel element. For fast and accurate detection of escaped coated fuel particles, X-ray DR (digital radiography) imaging with a step-by-step circular scanning trajectory was adopted for Chinese 10 MW HTGRs. The scanning parameters dominating the volume of the blind zones were optimized to ensure the missing detection of the escaped coated fuel particles is as low as possible. We proposed a dynamic calibration method for tracking the projection of the fuel-free zone accurately, instead of using a fuel-free zone mask of fixed size and position. After the projection data in the fuel-free zone were extracted, image and graphic processing methods were combined for automatic recognition of escaped coated fuel particles, and some practical inspection results were presented.

Near term test plan using HTTR (high temperature engineering test reactor)

JAEA has carried out research and development to establish the technical basis of high temperature gas cooled reactors (HTGRs) using HTTR. In order to connect hydrogen production system to HTTR, it is necessary to ensure the stability of plant dynamics when the thermal-load of the system is lost. Thermal-load fluctuation test is planned to demonstrate the stable reactor dynamics and to gain the test data for validation of the plant dynamics code. It will be confirmed that the reactor become stable state during a part of removed heat at HTTR heat-sink is lost. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations for safety analysis, and changes with burnup because of variance of fuel compositions. Measurement of temperature coefficient of reactivity has been conducted by HTTR to confirm the validity of the calculated temperature coefficient of reactivity. A loss of forced cooling (LOFC) test using HTTR has been carried out to verify the inherent safety of HTGR under the condition of loss of forced cooling while the reactor shut-down system disabled.

R&D plan for development of oxidation-resistant graphite and investigation of oxidation behavior of SiC coated fuel particle to enhance safety of HTGR

A new concept of the high temperature gas-cooled reactor (HTGR) is proposed as a challenge to assure no event sequences to the harmful release of radioactive materials even when the design extension conditions (DECs) occur by deterministic approach based on the inherent safety features of the HTGR. One of the DECs for the HTGR is the air/water ingress accidents. As the present status of the HTGR, such as the High Temperature Engineering Test Reactor (HTTR) in Japan Atomic Energy Agency (JAEA), graphite sleeves hold the fuel compact which consists of the SiC-TRISO (TRIstructural-ISOtropic) coated fuel particle, and make coolant flow channels. Then, the air/water ingress accident at the HTGR is prevented by additional measures (e.g. facility for suppression to air ingress). With regard to the core design, it is important to prevent recriticality accidents by keeping the geometry of the fuel rod which consists of the graphite sleeve, fuel compact and SiC-TRISO coated fuel particle, and by improving the oxidation resistance of the graphite when air/water ingress accidents occur. Therefore, it is planned to develop the oxidation-resistant graphite, which is coated with gradient SiC layer. It is also planned that the experimental identification of the condition to form the stable oxide layer (SiO2) for SiC layer on the oxidation-resistant graphite and on the SiC-TRISO fuel. This paper describes the R&D plan for un-irradiation and irradiation test under simulating air/water ingress accident condition to develop oxidation-resistant graphite and to investigate the oxidation behavior of SiC coated fuel particle.

Study of SiC/SiO2 oxidation-resistant coatings on matrix graphite for HTR fuel element

The improvement of the oxidation resistance of fuel element matrix graphite used in high temperature gas-cooled reactor (HTR) is indispensable for the further improvement of the safety of HTR. The stability of SiC coatings and SiC/SiO2 multilayer coatings in helium with little impurities is analyzed on the basis of thermodynamic calculation and it is found that SiC/SiO2 multilayered coatings are suitable oxidation resistant materials for graphite used in HTR fuel elements because they are stable both at the normal operation conditions and accidental conditions of HTR. SiC/SiO2 multilayer coatings were synthesized on the surface of fuel element matrix graphite of HTR by combination of vapor reaction diffusion, slurry method and high temperature oxidation in air; the oxidation behavior of SiC/SiO2 coated fuel element matrix graphite of HTR in different atmosphere are investigated and it is shown that SiC/SiO2 coatings can keep their stability under different experiment conditions which graphite used in HTR may encounter and greatly improve the oxidation resistance of fuel element matrix graphite of HTR.

A Thermohydrodynamic Analysis of Dry Gas Seals for High-Temperature Gas-Cooled Reactor

A comprehensive analysis method is proposed to resolve the problem of simulating the complex thermoflow with two kinds of distinct characteristic lengths in a dry gas seal. A conjugated simulation of the complicated heat transfer and the gas film flow is carried out by using the commercial computational fluid dynamics (CFD) software CFX. By using the proposed method, three-dimensional velocity and pressure fields in the gas film flow and the temperature distribution within the sealing rings are investigated for three kinds of film thickness, respectively. A comparison of thermohydrodynamic characteristics of the dry gas seal is conducted between the sealed gas of air and helium. The latter one is used in a helium compressor for a high-temperature gas-cooled reactor (HTGR). From comparisons and discussions of a series of simulation results, it will be found that the comprehensive proposal is effective and simulation results are reasonable. Even under a hypothetical accidental condition, the maximum temperature rise in the dry gas seal is within the acceptable range of HTGR safety requirements.

Numerical simulation of three-dimensional thermal-hydraulic behavior for HTTR (High Temperature Engineering Test Reactor)

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are now in progress in order to verify the inherent safety features and to improve safety designing and analysis technologies for future HTGR (high temperature gas-cooled reactor). Coolant flow reduction test is one of the safety demonstration tests for the purpose of demonstration of inherent HTGR safety features in the case that coolant flow is reduced by tripping of helium gas circulators. If reactor core element temperature and core internal structure temperature during abnormal events are estimated by numerical simulation with high-accuracy, developed numerical simulation method can be applied to future HTGR designing efficiently. In the present research, three-dimensional in- and ex-vessel thermal-hydraulic calculations for the HTTR are performed with a commercially available thermal-hydraulic analysis code “STAR-CD ®” with finite volume method. The calculations are performed for normal operation and coolant flow reduction tests of the HTTR. Then calculated temperatures are compared with measured ones obtained in normal operation and coolant flow reduction test. As the result, calculated temperatures are good agreement with measured ones in normal operation and coolant flow reduction test.

Post-test analysis of helium circulator trip without scram at 3 MW power level on the HTR-10

Safety demonstration tests were conducted on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the transient data of reactor core and primary cooling system for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 3 MW power level on October 15, 2003. This paper simulates and analyzes the power transient and the thermal response of the reactor during the test by using the THERMIX code. The analytical results are compared with the test data for validation of the code. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shut down after the stop of the helium circulator; the subsequent phenomena such as the recriticality and power oscillations are also studied. During the test a natural circulation loop of helium is established in the core and the other coolant channels and its consequent thermal response such as the temperature redistribution is investigated. In addition, temperatures of the measuring points in the reactor internals are calculated and compared with the measured values. Satisfactory agreements obtained from the comparison demonstrate the basic applicability and reasonability of the THERMIX code for simulating and analyzing the helium circulator trip ATWS test. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature during the test is always lower than 1600 °C which is the limited value for the HTGR.

A Study of Air Ingress and Its Prevention in HTGR

A rupture of the primary piping in the helium-cooled and graphite-moderated high-temperature gas-cooled reactor (HTGR) represents a design-basis event that should not result in significant safety consequences. In such a loss-of-coolant event, the reactor would be shut down inherently, and the decay heat would be removed passively with the ultimate reactor temperature rise being less than the design limit. Still, an important concern for reactor safety continues to be graphite oxidation damage to the fuel and core should a major air ingress take place through the breached primary pressure boundary. Two major cases of air ingress are studied. The first case results from the rupture of a control rod or refuel access standpipe atop the reactor pressure vessel (RPV). To rule out the possibility of such a standpipe rupture, a design change is proposed in the vessel top structure. The feasibility of the modified vessel local structure is evaluated. The second case of air ingress results from the rupture of one or more main coolant pipes on the lower body of the RPV. Experiment and analysis are performed to understand the multiphased air ingress phenomena in the depressurized reactor. Accordingly, a new passive mechanism of sustained counter air diffusion is proposed and shown to be effective in preventing major air ingress through natural circulation in the reactor. The results of the present study are expected to enhance the HTGR safety and economics.

ICONE15-10158 CORE DYNAMICS ANALYSIS FOR REACTIVITY INSERTION AND LOSS OF COOLANT FLOW TESTS USING THE HTTR

The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950℃ (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed. The reactor power decreases to decay heat level from the maximum reactor power of 30 MW due to the negative reactivity feedback effect of the core. Although the reactor power becomes critical again about five hours later, the peak power value is merely 2 MW. It was confirmed that by using the developed code, it is possible to not only analyze the reactor core dynamics but also simulate the core dynamics during the abnormal events postulated in the HTGR safety analysis.

Safety demonstration tests using high temperature engineering test reactor

Safety demonstration tests using the high temperature engineering test reactor (HTTR) are conducted for demonstrating inherent safety features of high temperature gas-cooled reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005 and the second phase tests will be carried out from FY 2006.

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).

HTGR plant safety design bases

The high temperature gas-cooled reactor (HTGR) has inherent and design safety features that are sifnificant and unique, requiring a number of safety criteria and approaches that differ markedly from other reactor types. This paper briefly reviews the design of HTGR plants that have been built and are being offered in the United States. It then reviews the safety considerations involved in the design of the plants being offered. The unique features, their development, and their effects on safety criteria are described. The design bases of the prestressed concrete reactor vessel (PCRV) are given particular attention. Operating characteristics of the HTGR and plant response to transient conditions are discussed. The design-basis depressurization accident evolution and related HTGR safety requirements are discussed. Characteristics of the HTGR with respect to technical specifications are discussed, with particular emphasis on the PCRV and the core safety limit.