Temperature Gas-cooled Reactor-Pebble Bed Research Articles

Estimation of transuranic isotopes in irradiated fuel elements for pebble bed HTR on activity vector

Due to the unknown power history of fuel elements in a pebble-bed High-Temperature Reactor, the traditional method of calculating transuranic isotopes through forward burnup calculations using initial loading data is not applicable.This paper presents a novel method for estimating transuranic isotopes within fuel elements in pebble bed HTR. It involves the power history inversion model using activity vector detectable by the online Burnup Measurement System of High Temperature Gas-cooled Reactor Pebble bed Module to predict the power history of the fuel elements. Subsequently, transuranic isotopes are calculated through a forward burnup process.To check the performance of the burnup equation inverse model. Nuclear Inventory Tool were used to generate database that provides known activity vector as target and corresponding power history of a fuel element. Results show that the inverse model leads to a more stable convergence, effectively reconstructing the complex power history. The integrated power estimation error remains below 0.2%, with prediction errors for ten transuranic isotopes of interest below 1.5%.

Experimental analysis of a large range of the load capacity for the axial AMB in the HTR-PM

An axial active magnetic bearing (AMB) is equipped to provide axial support of the vertical rotor of the helium circulator in the High Temperature Gas-cooled Reactor Pebble Bed Module (HTR-PM). The axial AMB is required to support the rotor and the aerodynamic load. High-strength alloy steel is selected as the material of the thrust disk. Maximum load capacity of the axial AMB in the HTR-PM helium circulator is analyzed with the consideration of the rotating velocity and the material strength. Also, experiments are conducted based on a 1:1 prototype to reveal a large range of the load capacity for the axial AMB. The dynamic electromagnetic force corresponding to the allowable stress, the rotational velocity, and other parameters of the axial AMB, such as B-R Curve, resistance and inductance, are revealed. Effect of the heat treatment to the load capacity is also discussed. The results will provide some theoretical and engineering reference to the design and the type selection of the axial AMB in the rotor system of the HTGR (High Temperature Gas-cooled Reactor) energy conversion system.

Steady-state thermal fluids analysis for the HTR-PM equilibrium core

The high temperature gas-cooled reactor-pebble bed module (HTR-PM), designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, is a demonstration project of nuclear power plant with modular high temperature reactors (HTRs) in China. So far the detailed engineering design of the HTR-PM has been completed, and now the plant is at its commissioning stage. In the design process of the reactor, the thermal hydraulics design is of extreme importance for the evaluation of the reactor performance under various operation conditions. In this study, the steady-state thermal fluids analysis is performed for the HTR-PM equilibrium core. The analysis methods are introduced and the calculating models are presented. The key thermal-fluid parameters under different power levels are obtained and evaluated, including the temperature field of the reactor, maximum temperatures of the fuel and other components, pressure loss of the core and the primary circuit, and so on. As the most important safety-related parameter, the maximum fuel temperature is much lower than its safety limit for normal operation. The analysis results show that the thermal hydraulics design provides sufficient cooling capability to the core and thus ensures the reactor safety under normal operation.

Rapid accident source term estimation (RASTE) for nuclear emergency response in high temperature gas cooled reactor

Rapid accident source term estimation (RASTE) is one of the key functions required by nuclear power plant (NPP) emergency. The results of RASTE are provided to the nuclear decision-making system in emergency, response to ranking accident by INES table and determining suitable protection. This paper proposes the RASTE system for high temperature gas-cooled reactor (HTR) to estimate accident source term based on the simplified methodology used in safety analysis report (SAR) and more realistic operation condition before and in an accident. RASTE simplifies the expression for normal radioactivity in the primary loop, re-suspended radioactivity, additional radioactivity release during the heating process, graphite dust generation, desorbed radioactivity from corrosion of graphite matrix and radioactivity released from failed coated fuel particle. Modifications are made to correct the different ingress water amounts and different fuel burnup. Some unclearly known parameters and processes are estimated from the operation experience as well as accident analysis by conservation rule. The integrated code of fundamental framework and user interface are also developed. This paper takes the High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM), a demonstration project being constructed at Shidao Bay, Shandong Province, China, as the study case to discuss the applicability of RASTE. Considering three typical HTR-PM design basis accidents (DBAs), this paper selects P2, P4 as the representatives of LOCA accidents, selects SG6 and SG91 as the representatives of ingress accidents and selects T2 as the representative of transient accidents. Accuracy verification is made by comparing the environmental release estimated by RASTE to that in SAR. These compared targets include nuclide groups of inert gas, iodine, and aerosols, and key nuclides of Xe-133, I-131, Cs-134. Calculation results show that the methodology presented and RASTE is fit to estimate accident of HTR-PM. The development of RASTE system will contribute to form adequate capability of rapid emergency response for HTR nuclear power plant.

Preliminary phenomena identification and ranking tables on the subject of the High Temperature Gas-cooled Reactor-Pebble Bed Module thermal fluids and accident analysis

The High Temperature Gas-cooled Reactor-Pebble Bed Module (HTR-PM) is a demonstration power plant designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University of China. In conformance with China’s nuclear safety requirements related to the development, assessment, application and quality assurance of computer software used in the safety analysis of nuclear power plants (NPPs), a phenomena identification and ranking table (PIRT) process was conducted on the subject of the HTR-PM thermal fluids and accident analysis by an expert panel sponsored by INET, on the basis of the latest HTR-PM design scheme. The PIRT objective is to identify safety-relevant phenomena that may occur during normal operation and under accident conditions, and to assess research activities pertaining to code development and assessment, such as the code validation in INET.In the PIRT process, safety-related phenomena were identified by the expert panel for the following HTR-PM scenarios: normal operation (NO), pressurized loss of forced cooling (PLOFC), depressurized loss of forced cooling (DLOFC), air ingress following DLOFC, steam/water ingress and anticipated transients without scram (ATWS). Subsequently, the relative importance of each phenomenon was evaluated according to its effect or influence on the figures of merit established by the expert panel, such as the fuel failure fraction and the maximum fuel temperature. In addition, each identified phenomenon was assigned a knowledge level rank by the expert panel, indicating the prediction capability for this phenomenon based on existing analytical tools and test data. What is more, all of the PIRT elements were well documented by a special report, including PIRT issues, objectives, descriptions of the HTR-PM plant and accident scenarios, evaluation criteria, current knowledge base, ranking scales of relative importance and knowledge level, a set of PIRTs, and expert discussions.

The influence of nuclear graphite oxidation on air ingress accident of HTR-PM

The high temperature gas-cooled reactor pebble bed module (HTR-PM) has wide potential applications in the development of advanced nuclear energy. The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct accident is a kind of beyond design basis accident of HTR-PM, which will result in air ingress into the reactor core. Even though the rare probability of this accident, it still has got a lot of concerns in the safety analyses of HTR-PM, due to the possible severe consequences of the air ingress accident, which would cause oxidation of the graphite material in the reactor. The oxidation of matrix graphite of spherical fuel elements may lead to the exposure of the TRISO fuel particles and the increase of the radioactivity release. The oxidation of the bottom reflector may cause the decrease of mechanical strength of the nuclear grade graphite. Different kinds of nuclear graphite of the bottom reflector will influence the consequences of the DEGB of the horizontal coaxial gas duct accident. This paper mainly analyzed the influence of the nuclear grade graphite used as the bottom reflector on the consequences of this accident by using the system analysis TINTE code.

Theoretical Analysis of Effective Thermal Conductivity for the Chinese HTR-PM Heat Transfer Test Facility

The Chinese high temperature gas-cooled reactor pebble bed module (HTR-PM) demonstration project has attracted increasing attention. In order to support the project, a large-scale heat transfer test facility has been constructed for pebble bed effective thermal conductivity measurement over the whole temperature range (0~1600 °C). Based on different heat transfer mechanisms in the randomly packed pebble bed, three different types of effective thermal conductivity have been theoretically evaluated. A prediction of the total effective thermal conductivity of the pebble bed over the whole temperature range is provided for the optimization of the test facility and guidance of further experiments.

Design and Development Framework of Safety-Critical Software in HTR-PM

With the development of information technology, the instrumentation and control system of nuclear power plant nowadays rely heavily on the massive and complex software to ensure the safe and efficient operation of the power plant. The improvement of the software design and development for the safety systems has been a research focus for its decisive impact on the nuclear safety. The framework of the software design and development for reactor protection system in High Temperature Gas-Cooled Reactor-Pebble bed Module was introduced in this paper. Firstly, during the design period, in addition to multichannel redundancy, grouping of protection variables and diverse 2-out-of-4 logics were adopted by different subsystems of each channel in case of common cause failure. Then a series of development characteristics together with strict software verification and validation were performed. Thirdly, during the software test period, an improved software reliability growth model based on the Goel-Okumoto model according to the analysis of fault severity was proposed to help in estimating the reliability of the software product and identifying the software release time.

An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor

An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor (HTGR) was developed and experimentally verified using the 10MW High-Temperature Gas-cooled Reactor-test module (HTR-10). Compared with the previous process, an auxiliary molecular sieve bed was added in helium purification regeneration system and new operation process was proposed to collect tritium-containing waste water. In this paper, the optimized process and verification experiment were presented in detail. In commissioning experiment of the improved HTR-10, a large quantity of high-dose tritium-containing waste water was successfully collected in the water separator of helium purification regeneration system, with the specific activity being 6.1×109Bq/L. The verification experiment confirms that the optimized process is effective and reliable for the demonstration plant design of High Temperature Gas-cooled Reactor-Pebble bed module (HTR-PM) to avoid the large emission of detrimentally radioactive waste water to the environment.

Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited from PSA is indeed the systematic way which PSA follows. This is more important especially for the new reactors.