Hot Isostatic Pressing (HIP) of type 316L stainless steel powder has been an established manufacturing practice for more than twenty-five years in the oil and gas sector and more recently in the naval defence sector. To demonstrate the capability of the powder metallurgy HIP (PM/HIP) for nuclear power applications a systematic study of 316L commercial powder production, encapsulation/consolidation providers and selected HIP parameters was undertaken by the Nuclear AMRC in collaboration with the Electric Power Research Institute (EPRI). In the study, the 316L powder specification limited the oxygen content of the powder to under 130 parts per million (ppm), which reflects the improvements that commercial powder suppliers have been making over the past decade to ensure greater powder cleanliness. The test programme assessed powder supply, HIP service provider and HIP sustain time. Excellent test results were achieved across the full range of variables studied with all billets meeting the specification requirements of ASTM A988 and additional requirements imposed based on nuclear manufacturing standards. Significantly, the study demonstrated the robustness of the PM/HIP supply chain, as material produced via differing HIP service providers resulted in very consistent material properties across the destructive test programme. Furthermore, no significant difference in material properties were noted for material HIP’ed between 2–8 hours hold time, suggesting that the HIP process window is large. Both these results are significant from an end-user standpoint as they highlight the uniformity of the process through the full manufacturing cycle from powder procurement to destructive testing. Despite all material passing specification requirements, some property variation was noted for differing powder suppliers. Considering the systematic approach, this was attributed to powder composition, with both low oxygen and high nitrogen contents contributing to improvements in Charpy impact strength and tensile strength respectively.

Hydrogen combustion or detonation happened in the containment within the process of the small reactor severe accident may threaten the integrity of the containment. In this paper, based on systemic design of the Small Modular Reactor (SMR) surrounded by the steel containment, an innovatory combustible gas control strategy which using the passive containment cooling system (PCCS) and passive autocatalytic recombiners (PARs) is made to control the hydrogen risk in the small steel containment. A severe accident hydrogen risk analysis model is built by the integrative severe accident analysis program MELCOR, the validity of the strategy is analyzed at a typical severe accident. With this understanding, a three-dimensional computed fluid dynamics hydrogen behavior analysis model of the small steel containment is established by GASFLOW code, and the gas distribution non-uniformity in the containment is analyzed. The result shows that the steam condensation process in the containment could be slowed down by controlling the action of PCCS, and the steam concentration in the containment could be in the range of high level, while the oxygen concentration could be in the range of low level. If the PARs were added, the PARs could consume the hydrogen and oxygen in the containment sustainedly. The containment atmosphere could be in an inerted condition during the accident process, even though the hydrogen concentration in the containment is high. The gas distribution non-uniformity analysis result shows that oxygen concentration was low in the extent of high hydrogen concentration and high steam concentration, the steam, oxygen and hydrogen distribution non-uniformity would not affect the inerted atmosphere of containment.

The influence of gas introduction on the critical safety of the nuclear fuel system under the condition of cold condition, given reactor material and geometry structure is studied. Refer to bubble effect test experiment on nuclear critical safety test device (YSR) and considering solid-liquid two-phase nuclear fuel system with uranyl nitrate solution - uranium dioxide fuel element as the experimental platform, the dynamic process of the real behavior of bubbles in uranyl nitrate solution has been simulated in the quasi-static way by replacing bubble generator with aluminous bubble simulation elements. Bubble effect is the reactivity change caused by the change of volume of solution, neutron leakage and absorption property in the nuclear fuel system due to the bubbles generated in the solution. In the dissolving process of spent fuel, oxygen or nitrogen are usually added to accelerate the dissolution of fuel element shear section, and some other bubble production are also caused by the heat released during the dissolution process. Here, the bubble production caused by the heat is omitted and only artificial gas introduction is considered in my study. When there are enough bubbles in the uranium solution system, the volume of the solution will increase, which will inevitably influence the absorption and leakage property of the neutrons, and further influence the reactivity of the nuclear fuel system. The corresponding relationship between the bubble-intake rate and the bubble equivalent diameter, arising velocity and bubble share is determined through fluid dynamics modeling to manufacture the aluminous bubble simulation elements. The theoretical calculation by MONK9A and the critical experimental measurements are also compared and analyzed in this paper. The results showed that the reactivity caused by bubbles was negative, and the greater the intake rate, the greater the negative effect. Meanwhile the theoretical calculated value was in good agreement with the experimental value and the maximum deviation was 63.4 pcm.

Authors are developing an experimental technology to realize experiments simulating Severe Accident (SA) conditions using simulant fuel material (ZrO 2 with slight addition of MgO for stabilization) that would contribute not only to Fukushima Daiichi (1F) decommissioning but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of the accident progression behavior. Based on the results of the prototype test, improvement of plasma heating technology was conducted. The Core Material Melting and Relocation (CMMR)-1/-2 experiments were carried out in 2017 with the large-scale simulated fuel assembly (1 m × 0.3 mϕ) applying the improved technology (higher heating power and controlled oxygen concentration). In these two tests, heating history was different resulting basically in similar physical responses with more pronounced material melting and relocation in the CMMR-2 experiment. The CMMR-2 experiment is selected here from the viewpoint of establishing an experimental technology. The CMMR-2 experiment adopted 30-min heating period, the power was increased up to a level so that a large temperature gradient (> 2,000 K/m) expected at the lower part of the core in the actual 1F accident conditions. Most of the control blade and the channel box migrated from the original position. After the heating, the simulated fuel assembly was measured by the X-ray Computed Tomography (CT) technology and by Electron Probe Micro Analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective in terms of applicability of the non-transfer type plasma heating technology to the SA experimental study was obtained.

A large amount of hydrogen is generated by the metal water-reaction in the Primary containment vessel (PCV) of light water reactors in the severe accident (SA). In the present accident management for boiling water reactor (BWR), vent of mixing gas with filtered vent is regarded as the most likely method that prevents the PCV overpressure. However, it is difficult to vent in early stage of SA because of high radioactive dose. Then we have been developing the hydrogen treatment system to prevent excessive pressure without PCV vent. In focusing on the oxidation-reduction reaction of metal oxides (MOs) with high reaction rate, we have been studying hydrogen treatment system using MOs as effective device under oxygen deficit conditions like PCV of BWR. In the previous studies, we evaluated the hydrogen treatment rate using a couple of MOs, and confirmed that CuO, Co 3 O 4 , and MnO 2 were effective for the hydrogen oxidation under the oxygen-free condition. We also found that granules of these three MOs could achieve the goal of hydrogen treatment rate with reactor of hydrogen treatment system. We predicted that the performance of MOs decreased with exposure to steam and fission products (FPs) in the PCV during the hydrogen treatment, and investigated their influence. The objective of the present research is to investigate how the steam and FPs, which is supposed to be a reaction-inhibiting-factor, influence hydrogen treatment rate. Then, we conducted hydrogen treatment experiments using a fixed bed reactor with MOs layer. As the results, we confirmed that the hydrogen treatment rate of MnO 2 decrease from 70 g/s/m 3 to 15 g/s/m 3 when partial pressure of vapor went above 0.1 MPa-abs, though, that of CuO didn’t depend on the partial pressure of vapor and sustain the same rate about 40 g/s/m 3 . We also confirmed that the hydrogen treatment rate was decreased with the consumption of granulated MOs faster than our expectation estimated with unreacted-core model*. We also estimated that CsI selected as typical FPs could not affect the hydrogen treatment rate of CuO. From these results, we have evaluated the reaction rate equation including the steam influence in CuO, which could estimate the hydrogen treatment rate of reactor unit. *Gas reacts only on the surface of solid and generates shell of products around reactants core. The core shrinks with reaction.

A large amount of hydrogen is generated by the metal water-reaction in the Primary Containment Vessel (PCV) of Light Water Reactors in the severe accident. Then we have been developing the Hydrogen Treatment System to prevent excessive pressure without PCV vent. By focusing on the oxidation-reduction reaction of metal oxides with high reaction rate, we have been studying hydrogen treatment system using metal oxides as effective device under oxygen deficit conditions of PCV of Boiling Water Reactor (BWR). We have been considering a hydrogen treatment unit with a lot of pipes in which metal oxides are filled. We have already investigated experimentally the basic trend of metal oxides temperature and gas concentration as well as hydrogen processing rate dependency on gas temperature and concentration in the absence of steam. However, the influence of steam on hydrogen processing characteristics has not been clarified yet. The objective of the present research is to investigate how the steam, which is supposed to be a reaction-inhibiting-factor, affects hydrogen processing characteristics of the hydrogen treatment unit. We conducted experiments using a test section of one pipe simulating a part of the hydrogen treatment pipes. The granulated CuO, which is a candidate material for the actual system was used. The hydrogen concentration of 10 wt% at the inlet of the pipes was decreased to 0 wt% at the outlet even in high steam concentration conditions (30–50 wt%) when the gas temperature was 250 °C, therefore, it was confirmed that hydrogen was treated with high processing rate under steam circumstances. It was also found that cumulative amount of treated hydrogen was strongly correlated with temperature and relative humidity. We have been developing the thermal-chemical model of hydrogen treatment unit. The prediction margin of error was decreased to 30 % from over 100 % by improving degradation model based on the experiment results, therefore, it reached the practical level.

With the development of small modular reactors, the hydrogen risk reducing technology cannot be ignored. Special safety facilities of small modular reactor (SMR) are investigated and studied, and a serious accident analysis program model for SMR is established. The combination of Pre-inerting and hydrogen recombination was used to control the hydrogen risk. The effectiveness of the hydrogen control system is analyzed by using the GASFLOW program. The results show that the volume fraction of hydrogen in the containment dome is higher than that in the other parts of the containment during the calculation. Because of the small size and tight internal structure, hydrogen accumulates in the narrow channel, which increases the hydrogen concentration in the local channel. Inerting reduces the concentration of oxygen in the containment and effectively controls the possibility of flame acceleration and blasting transition in high hydrogen concentration regions.

A FLiNaK high temperature test loop, which was designed to support the Thorium Molten Salt Reactor (TMSR) program, was constructed in 2012 and is the largest engineering-scale fluoride loop in the world. The loop is built of Hastelloy C276 and is capable of operating at the flow rate up to 25m 3 /h and at the temperature up to 650°C. It consists of an overhung impeller sump-type centrifugal pump, an electric heater, a heat exchanger, a freeze valve and a mechanical one, a storage tank, etc. Salt purification was conducted in batch mode before it was transferred to and then stored in the storage tank. The facility was upgraded in three ways last year, with aims of testing a 30kW electric heater and supporting the heat transfer experiment in heat exchanger. Firstly, an original 100kW electric heater was replaced with a 335kW one to compensate the overlarge heat loss in the radiator. A pressure transmitter was subsequently installed in the inlet pipe of this updated heater. Finally, a new 30kW electric heater was installed between the pump and radiator, the purpose of which was to verify the core’s convective heat transfer behavior of a simulator design of TMSR. Immediately after these above works, shakedown test of the loop was carried out step by step. At first the storage tank was gradually preheated to 500°C so as to melt the frozen salt. Afterwards, in order to make the operation of transferring salt from storage tank to loop achievable, the loop system was also preheated to a relatively higher temperature 530°C. Since the nickel-base alloy can be severely corroded by the FLiNaK salt once the moisture and oxygen concentration is high, vacuum pumping and argon purging of the entire system were alternatively performed throughout the preheating process, with the effect of controlling them to be lower than 100ppm. Once the salt was transferred into the loop, the pump was immediately put into service. At the very beginning of operation process, it was found that flow rate in the main piping could not be precisely measured by the ultrasonic flow meter. Ten days later, the pump’s dry running gas seal was out of order. As a result, the loop had to be closed down to resolve these issues.

Water chemistry plays an important part in maintaining corrosion resistance in water transport systems throughout nuclear power plants (NPP’s). Small changes in liquid chemistry such as pH, borate concentration, or build-up of crud in reactor cooling water can result in rapid degradation or damage to components and lead to unexpected failures. The Chemical and Volume Control System (CVCS) and Reactor Water Cleanup System (RWCU) are responsible for maintaining these parameters at appropriate levels, and so failure of either of these systems can result in unnecessary stresses on many other reactor systems due to resulting transients. While the major components of these systems all have sufficient redundancy to prevent major accidents, failure of components in these systems can result in failure of other redundant components and affect plant safety [1]. The CVCS and RWCU systems have experienced aging related degradations and failures in the past, and although they have not affected the system’s emergency functions, they have resulted in unnecessary actuation of related systems, and reactor shutdowns [1]. Reactor shutdowns can result in large changes in reactor coolant chemistry such as oxygen and borate concentration transients, and the build-up of corrosion products which can’t be as easily removed during periods of reactor shutdown [2]. In the following analysis of Component Operational Experience Degradation and Ageing Program (CODAP) experience data; causes, impacts, and preventative actions as recorded in CODAP are examined for degradation events which took place in the CVCS and RWCU, of PWRs and BWRs, respectively. The analysis will demonstrate the usefulness of CODAP in examining reactor component failure trends, as well as discuss insights on improvement for the program.

As for the decommissioning of Fukushima Dai-ichi nuclear power plant (1F), a long-term waste storage container with high safety is requested to store radioactive materials such as fuel debris for a long period of time. Since hydrogen is generated by radioactive decomposition of water, it is important to keep the concentration of hydrogen gas below the explosion limit in order to ensure the safety of the container. Then, use of passive autocatalytic recombiner (PAR) was proposed to reduce the hydrogen concentration. PAR is installed in the container. In order to experimentally confirm the reduction of hydrogen concentration by PAR and hydrogen behavior in the container, an experimental apparatus consisting of a small-scale modeled container and a hydrogen supply system was provided. Preliminary experiments were begun for confirming fundamental performance of the experimental apparatus under the conditions that PAR and simulated fuel debris are not installed in the container. Moreover, the hydrogen behavior in the container was analyzed numerically. In addition, the steam behavior generated by the chemical reaction of hydrogen and oxygen by PAR was also predicted. This paper describes both results of the preliminary experiments and numerical simulations. The experimental results showed that the hydrogen behavior can be predicted using the temperature distributions in the container. The analysis results clarified the controlling factors on the hydrogen behavior and the steam distribution in the container by PAR.

In the reactor core of a pebble bed HTGR, fuel pebbles are in random arrangement. The coolant flow passages are complicated. Under the air and water ingress accident conditions, the partially enriched oxidizing gas corrodes the fuel pebbles by oxidation reaction. How the oxidizing gases (oxygen and water vapor) diffuse in the pebble bed will greatly influence the oxidation process. Analysis of the mass mixing is important for analyzing the fuel element graphite corrosion and safety of HTGR. In this paper, a three-dimensional simple cubic structure packed cells of pebble bed core model with surface contact method was established. The flow, and mass mixing processes were numerically and theoretically studied using numerical simulation and porous media methods. The effects of molecular diffusion, turbulent diffusion and mass dispersion on mass mixing effect were discussed respectively.

Accumulative test data indicates that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of materials used in the reactor coolant pressure boundary components to significantly reduce. EAF is used as the abbreviation of the environmentally assisted fatigue in the nuclear field. In 2007, NRC issued RG. 1.207. It was updated in 2014. And it requires that the effects of the light-water environment on the fatigue life reduction of metal components should be considered for new plants. And it suggests to use environmental correction factor (Fen) to account for EAF. Fen = Nair/Nwater (N is occurrences). NUREG/CR-6909 [1] presents the detail Fen calculation formula which includes the complicated influence of combined multi-parameters. Fen is a function of temperature, strain amplitude & rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depends on the experience of the primary pressure boundary piping transient operation, Fen vary during each transient. More uncertainty and confusion are raised during the application of the Fen method. In the research work involved in this article, first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depend on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multi-transient loading conditions. And the result is compared with that of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test matrix is needed to prove its practicability furthermore.

The hydrogen emission of zirconium hydride at high temperature is a challenging issue for many researchers. The hydrogen emission content of zirconium hydride pins should be evaluated to confirm the application feasibility. The comparison of theory analysis and experiment data indicated Richardson’s Law could offer a conservative result for calculating the hydrogen emission content of zirconium hydride pins at high temperature. Furthermore, the methods of preventing hydrogen loss should be developed for the purpose of extending the work temperature or time. The results showed a ZrO 2 layer prepared for zirconium hydride could not prevent hydrogen loss after exposure at 923K in an inert environment and ZrO 2 transformed into Zr 3 O gradually due to the opposite movement of hydrogen and oxygen. Finally, a further improvement to prevent hydrogen loss was developed. The zirconium hydride with a ZrO 2 layer in the cladding of He+CO 2 exhibited no significant reduction of hydrogen content. It is helpful to prevent the hydrogen loss by increasing the oxygen potential on the outside of ZrO 2 layer.

The oxide films formed on non-charged and hydrogen-charged Alloy 690 specimens exposed to 290 °C pure water environment with different dissolved oxygen concentrations were characterized. It was found that the oxide film formed on the hydrogen-charged specimen was thicker than that on the noncharged specimen. Local protruding oxides were observed on the hydrogen-charged Alloy 690 but not on the non-charged specimen. The Ni and Fe contents of local protrusion on the hydrogen-charged specimen were higher than those on the noncharged specimen.

The complex microstructures of stainless steel (SS) cladding on low alloy steel (LAS) joint and the corrosion behavior in high temperature water environments were examined. Scanning electron microscopy, focused ion beam, transmission electron microscope and Raman spectroscopy were used. The heat-affected zone (HAZ) in the low alloy steel was mainly comprised of pearlite and ferrite, while the HAZ in the stainless steel was mainly comprised of austenite and ferrite. The HAZ in the low alloy steel was divided into overheated crystal region, complete recrystallization region and incompletely recrystallization region. A decarburization zone in the low alloy steel side and a carbon-enriched zone in the stainless steel side were identified. M 23 C 6 and M 7 C 3 precipitation were observed mainly in carbon-enriched zone. The surfaces of the weld after corrosion in simulated PWR primary water environment at 290 °C were significantly affected by dissolved oxygen. In aerated solution, the oxide film on A508III steel was mainly γ-Fe 2 O 3 , which becomes spinel oxide on the 309L/308L cladding. In deaerated solution, the oxide film was mainly Fe 3 O 4 on A508III steel, which was spinel oxide on the 309L/308L cladding.

Uranium dioxide (UO 2 ) is the typical fuel that is used in nuclear fission reactor, fission gas are produced during and after the reactor operation, and the fission gas have a significant impact on the performance of UO 2 in reactor. In this paper, we investigated the effects of the fission gas on the performance of UO 2 by using the first-principles calculation method based on the density functional theory. The results are that, the volume of UO 2 increased when there is a fission gas atom enter in UO 2 supercell; fission gas prefer to occupy the octahedral interstitial site over the uranium vacancy site and the oxygen vacancy site, and the oxygen vacancy site is the most difficult occupied site due to the formation of an oxygen vacancy is more difficult than that of the uranium vacancy; our results of the UO 2 elastic constants are in good agreement with other simulation results and experimental data, and the fission gas atoms make the ductility of UO 2 decreased. Our works may shed some light on the development of the UO 2 fuel and the spent fuel reprocessing.

This work presents first results of the study on the influence of the LBE oxygen concentration on the initiation of dissolution corrosion in 316L austenitic stainless steels. 316L steel specimens were exposed at 450 °C to static liquid LBE with controlled and constant oxygen concentration of 10 −5 , 10 −6 and 10 −7 mass% for 1000 hours. Corroded specimens were analysed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). Limited oxidation corrosion and no dissolution corrosion was observed in the specimens exposed to LBE containing 10 −5 and 10 −6 mass% oxygen, while dissolution corrosion with a maximum depth of 59 μm was found in the specimen exposed to LBE containing 10 −7 mass% oxygen.

JAEA (Japan Atomic Energy Agency) has conducted feasibility studies of the fuel and of the reactor core for the plutonium-burner HTGR (High Temperature Gas-cooled Reactor). The increase of the internal pressure, which is caused by generations of CO gas and stable noble gases, is considered to be the one of the major causes of TRISO (TRI-structural ISO-tropic) fuel failure at high burn-up. The CO gas is generated by the chemical reaction of the graphite making up the buffer layer with the free-oxygen released from the fuel kernel by fission. The stable noble gases, which are fission products, are also released from the fuel kernel. Although it is considered very difficult to suppress the increase of the partial pressure of the stable noble gases because of its chemically inert nature, the increase of the CO gas partial pressure can be suppressed by reducing the free-oxygen mole concentration using a chemical reaction. ZrC acts an oxygen getter, which reduces the free-oxygen generated with fission reaction. An increase of the CO gas partial pressure with burn-up in a TRISO fuel is expected to be suppressed by coating ZrC on a fuel kernel. A PuO 2 -YSZ (Yttria Stabilized Zirconia) fuel kernel with a ZrC coating, which enhances safety, security and safeguard, namely: 3S-TRISO fuel, was proposed to introduce to the plutonium-burner HTGR. In this study, the efficiency of the ZrC coating as the free-oxygen getter under a HTGR temperature condition was examined based on a thermochemical calculation. A preliminary feasibility study on the 3S-TRISO fuel that enables to attain a high burn-up around 500 GWd/t was also conducted focusing on a fuel failure caused by an increase of the internal pressure. Additionally, a preliminary nuclear analysis was conducted for the plutonium-burner HTGR with a fuel shuffling in the radial direction. As a result, the thermochemical calculation result showed that all the amount of the free-oxygen is captured by a thin ZrC coating under 1600°C condition. The plutonium-burner HTGR will be designed to suppress fuel temperature to be lower than 1600°C under severe accident conditions, and hence it was confirmed that coating ZrC on the fuel kernel is very effective method to suppress the internal pressure. The internal pressure the 3S-TRISO fuel at 500 GWd/t is calculated to be lower than 60 MPa, which allows to prevent the fuel failure, and hence the feasibility of the 3S-TRISO fuel was also confirmed. Additionally, the results of the whole core burn-up calculations showed that the fuel shuffling in the radial direction allows to achieve the high burn-up around 500 GWd/t. It also showed that the temperature coefficient of reactivity is negative value during the rated power condition through the operation period.

The hydrogen treatment system has been developed in order to prevent the overpressure of the primary containment vessel (PCV) caused by a large amount of hydrogen generated by the metal-water reaction in severe accidents (SAs) of Light Water Reactor. In previous studies, we evaluated the hydrogen treatment rate using a couple of metal oxides, and confirmed that MnO 2 , CuO, and Co 3 O 4 were effective for the hydrogen oxidation under the oxygen-free condition, then we selected them as reactants[1]. Although the reactants were granulated with a diameter of 2 mm for application to the system, the hydrogen treatment rate has not been scarcely evaluated for the granulated MnO 2 which is expected to treat the hydrogen around 120 °C [2]. Thus, we made the diameter of the granulated MnO 2 smaller, and found that the hydrogen treatment was occurred by the granulated MnO 2 with a diameter below 1.0 mm. The granules with a diameter below 1.0 mm were also acceptable for the system from the point of view of decreasing the differential pressure (DP). Moreover, the experiments using a test section simulating a reactor of the system had been conducted under the hydrogen condition simulating typical condition of a SA, by loading the granulated CuO with a diameter of 2mm onto the granulated MnO 2 with a diameter of 1mm. As a result, the hydrogen treatment was markedly accelerated by supplying enough reaction heat from the granulated MnO 2 to the granulated CuO.

Extensively released hydrogen due to zirconium-water reaction during severe accidents in containment of pressurized water reactor raises explosion crisis. Since the containment is the last barrier for fatal irradiation species, efficient measures should be implemented to restrain the hydrogen. Hence, hydrogen elimination and monitor devices are wildly applied to address this issue. Detection of hydrogen once has been conducted by a traditional hydrogen concentration measuring system with sampling devices and hydrogen sensors that located outside the containment. This arrangement, however, is a compromise between actual requirements for hydrogen measuring device and absence of favorable hydrogen sensors which could be applied in the harsh environment under severe accidents. Most recently, R&D of hydrogen concentration measuring system with in-situ hydrogen sensor has attracted increasing attention. Mitsubishi Heavy Industries, Ltd is focusing their job on an electrochemical hydrogen sensor based on solid state electrolyte. Besides, AREVA has developed a system depending on thermal conductivity detector associated with catalytic combustion sensor which requires external power supply to heat the assembly. In PERIC, we have developed a hydrogen concentration measuring system with in-situ hydrogen sensor which can be set in accident confident area. The hydrogen sensor is originally based on catalytic recombination of hydrogen and oxygen. Generally, catalyst prepared using noble metals such as platinum and palladium is scientifically loaded in the hydrogen sensor to serve as hydrogen sensitive material. On the event of severe accidents, mixture of hydrogen and air can spontaneously diffuse into the hydrogen sensor, where, part of the mixture is involved in a chemical exothermic reaction on the catalyst to generate water and heat. Generally, an increased concentration of hydrogen will raise relatively higher reaction temperature of the hydrogen sensor. The hydrogen concentration related temperature of the hydrogen sensor is detected using steel armored thermocouple. Besides, environmental temperature and pressure in the containment are also acquired to assist calculation. All the data are transferred to a signal processing cabinet, which, performs the calculation and indication functions using programmable logic controller and digital display device, respectively. There is no organic material, mechanical moving and power consumption part in the hydrogen sensor and thermocouple. The system indicated reliable performance in simulated containment under condition of high temperature, pressure, steam, and etc. The hydrogen concentration measuring system illustrated excellent endurance to poisoning species such as iodine and aerosol. Furthermore, the hydrogen sensor also suggested high resistance to irradiation. The system can survive a severe earthquake, and its seismic certification toward to safety shutdown earthquake is class I. Over 80 systems so far have be applied in pressurized water reactor in China and or Pakistan. The latest model is designed according to requirements of CAP1400.