scholarly journals The R&D of HTR-STAC Program Package: Source Term Analysis Codes for Pebble-Bed High-Temperature Gas-Cooled Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-9 ◽  
Author(s):  
Hongyu Chen ◽  
Chuan Li ◽  
Haoyu Xing ◽  
Chao Fang
Keyword(s):  
High Temperature ◽  
Nuclear Reactor ◽  
Source Term ◽  
Safety Analysis ◽  
Operating Conditions ◽  
Primary Circuit ◽  
Pebble Bed ◽  
Water Reactors ◽  
Term Analysis ◽  
High Temperature Gas

Source term analysis is important in the design and safety analysis of advanced nuclear reactor and also provides a radiation safety analysis basis for Modular High-Temperature Gas-Cooled Reactor (HTR). High-Temperature Gas-Cooled Reactor-Pebble-bed Modules (HTR-PM) design by China is a typical Gen-IV and due to different safety concepts and systems, the implements of source term analysis in light water reactors are not entirely applicable to HTR-PM. To solve this problem, HTR-PM Source Term Analysis Code (HTR-STAC) has been developed and related V&V has been finished. HTR-STAC consists of five units, including LOOP (Primary Circuit Source Term Analysis Code), NORMAL (Normal Condition Airborne Source Term Analysis Code), ARCC (Accident Release Category Calculation code), CARBON (C-14 Source Term Analysis Code), and TRUM (Tritium Source Term Analysis Code). LOOP and NORMAL may be used as calculating primary circuit coolant radioactivity and the release of airborne radioactivity to the environment under normal operating conditions of HTR-PM, respectively. The code ARCC composed of several source term analysis programs in the different typical accidents scenario, including SGTR (Steam Generator Tube Rupture), LOCA (Loss of Coolant Accident), and the Transient Process, is compiled based on the results given by LOOP and NORMAL. CARBON and TRUM are developed to calculate the productions of C-14 and H-3 through a different mechanism. Furthermore, the V&V has been performed and show some positive results.

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

2010 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
Nodal Model ◽  
High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

scholarly journals Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

2017 ◽  
Vol 2017 ◽  
pp. 1-6 ◽  
Author(s):  
Xuegang Liu ◽  
Xin Huang ◽  
Feng Xie ◽  
Fuming Jia ◽  
Xiaogui Feng ◽  
...  
Keyword(s):  
High Temperature ◽  
Source Term ◽  
Pressurized Water Reactor ◽  
Sample Collection ◽  
Analytical Methodology ◽  
Pressurized Water ◽  
The Core ◽  
Irradiated Graphite ◽  
Term Analysis ◽  
High Temperature Gas

The high temperature gas-cooled reactor (HTGR) has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR) is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10) in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

Water-Ingress Accident of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2009 ◽  
Author(s):  
Zheng Yanhua ◽  
Shi Lei
Keyword(s):  
High Temperature ◽  
Safety Valve ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Water Ingress ◽  
Design Basis ◽  
Design Basis Accident ◽  
Water Gas ◽  
High Temperature Gas

Water-ingress accident, caused by the steam generator heating tube rupture of a high temperature gas-cooled reactor, will introduce a positive reactivity to lead the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Increase of the primary circuit pressure may result in the opening of the safety valve, which will cause the release of radioactive isotopes and flammable water gas. The analysis of such an important and particular accident is significant for verifying the inherent safety characteristics of the pebble-bed modular high temperature gas-cooled reactor. Based on the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the design basis accident of double-ended guillotine break of a heating tube has been analyzed by using TINTE, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature and primary loop pressure, the graphite corrosion inventory, the water gas releasing amount, as well as the natural convection influence under the condition of the failure of the blower flaps shut down, have been studied in detail. The calculation result of the design basis accident indicates that, the maximal possible water ingress amount is less than 600 kg and the maximal fuel temperature keeps far below the design limitation of 1620°C. The result also shows that the slight amount of graphite corrosion will not damage the reactor structure and the fuel element, and there is no potential explosive risk caused by the opening of the safety valve.

scholarly journals Double Heterogeneity Neutronic Simulation of HTR-10 for First Criticality and Full Power Initial Core with TORT-TD code

2021 ◽  
Vol 927 (1) ◽  
pp. 012037
Author(s):  
Daddy Setyawan
Keyword(s):  
High Temperature ◽  
Cross Section ◽  
Power Distribution ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
The Core ◽  
Full Power ◽  
Double Heterogeneity ◽  
High Temperature Gas

Abstract In order to support the verification and validation of computational methods and codes for the safety assessment of pebble bed High-Temperature Gas-cooled Reactors (HTGRs), the calculation of first criticality and full power initial core of the high-temperature pebble bed reactor 10 MWt (HTR-10) has been defined as one of the problems specified for both code-to-code and code-to-experiment benchmarking with a focus on neutronics. HTR-10 Experimental facility serves as the source of information for the currently designed high-temperature gas-cooled nuclear reactor. It is also desired to verify the existing codes against the data obtained in the facility. In HTR-10, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in the reactor consist of TRISO particles, which are embedded in the graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneity of this type of reactor. In this paper, the first criticality and the power distribution in full power initial core calculations of HTR-10 are used to demonstrate treatment of this double heterogeneity using TORT-TD and Serpent for cross-section generation. HTR-10 has unique characteristics in terms of the randomness in geometry, as in all pebble bed reactors. In this technique, the core structure is modeled by TORT-TD, and Serpent is used to provide the cross-section in a double heterogeneity approach. Results obtained by TORT-TD calculations are compared with available data. It is observed that TORT-TD calculation yield sufficiently accurate results in terms of initial criticality and power distribution in full power initial core of the HTR-10 reactor.

scholarly journals Analysis of Fission Products’ Release in Pebble-Bed High-Temperature Gas-Cooled Reactor Fuel Elements Using a Modified FRESCO II Numerical Model

2021 ◽  
Vol 2021 ◽  
pp. 1-8
Author(s):  
Chao Fang ◽  
Chuan Li ◽  
Jianzhu Cao ◽  
Ke Liu ◽  
Sheng Fang
Keyword(s):  
High Temperature ◽  
Source Term ◽  
Fission Products ◽  
Release Rates ◽  
Time Step ◽  
The Difference ◽  
Advanced Nuclear System ◽  
Fuel Elements ◽  
Term Analysis ◽  
High Temperature Gas

The radiation safety design and emergency analysis of an advanced nuclear system highly depends on the source term analysis results. In modular high-temperature gas-cooled reactors (HTGRs), the release rates of fission products (FPs) from fuel elements are the key issue of source term analysis. The FRESCO-II code has been established as a useful tool to simulate the accumulation and transport behaviors of FPs for many years. However, it has been found that the mathematical method of this code is not comprehensive, resulting in large errors for short-lived nuclides and large time step during calculations. In this study, we used the original model of TRISO particles and spherical fuel elements and provided a new method to amend the FRESCO-II code. The results show that, for long-lived radionuclides (Cs-137), the two methods are perfectly consistent with each other, while in the case of short-lived radionuclides (Cs-138), the difference can be more than 1%. Furthermore, the matrix method is used to solve the final release rates of FPs from fuel elements. The improved analysis code can also be applied to the source term analysis of other HTGRs.

scholarly journals ANALISIS INVENTORI REAKTOR DAYA EKSPERIMENTAL JENIS REAKTOR GAS TEMPERATUR TINGGI

2016 ◽  
Vol 22 (1) ◽  
Author(s):  
Sri Kuntjoro ◽  
Pande Made Udiyani
Keyword(s):  
High Temperature ◽  
Cross Section ◽  
Operating Conditions ◽  
Gas Temperature ◽  
Pebble Bed ◽  
Experimental Reactor ◽  
Group I ◽  
Reactor Type ◽  
Type Power ◽  
High Temperature Gas

ANALISIS INVENTORI REAKTOR DAYA EKSPERIMENTAL JENIS REAKTOR GAS TEMPERATUR TINGGI. Berkaitan dengan rencana Badan Tenaga Nuklir Nasional (BATAN) untuk mengoperasikan reaktor eksperimental jenis Reaktor Gas Temperatur Tinggi (RGTT), maka diperlukan analisis keselamatan terhadap reaktor terutama yang berkaitan dengan issue lingkungan. Analisis sebaran radionuklida dari reaktor ke lingkungan pada kondisi operasi normal atau abnormal diawali dengan estimasi sumber radionuklida di teras reaktor (inventori teras) berdasarkan pada tipe, daya, dan operasi reaktor. Tujuan penelitian adalah melakukan analisis inventori teras untuk disain Reaktor Daya Eksperimental (RDE) jenis reaktor gas temperature tinggi berdaya 10 MWt, 20 MWt dan 30 MWt. Analisis dilakukan menggunakan program ORIGEN2 berbasis pustaka penampang lintang pada temperatur tinggi. Perhitungan diawali dengan membuat modifikasi beberapa parameter pustaka tampang lintang berdasarkan temperatur rata-rata teras sebesar 5700 °C dan dilanjutkan dengan melakukan perhitungan inventori reaktor untuk reaktor RDE berdaya 10 MWt. Parameter utama reaktor RDE 10 MWt yang digunakan dalam perhitungan sama dengan parameter utama reaktor HTR-10. Setelah inventori reaktor RDE 10 MWt diperoleh, dilakukan perbandingan dengan hasil dari peneliti terdahulu. Berdasarkan kesesuaian hasil yang didapat dilakukan desain untuk reaktor RDE 20MWEt dan 30 MWt untuk memperoleh parameter utama reaktor tersebut berupa jumlah bahan bakar pebble bed di teras reaktor, tinggi dan diameter teras. Berdasarkan pareameter utama teras dilakukan perhitungan inventori teras RDE 20 MWt dan 30 MWt dengan metode yang sama dengan metode perhitungan pada RDE 10 MWt. Hasil yang diperoleh adalah inventori terbesar untuk reaktor RDE 10 MWt, 20 MWt dan 30 MWt secara berurutan untuk kelompok Kr adalah sekitar 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq untuk kelompok I sebesar 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq dan untuk kelompok Cs sebesar 2,20E+16 Bq, 2,40E+16 Bq dan 2,60E+16 Bq. Inventori teras selanjutnya akan digunakan untuk menghitung suku sumber dari reaktor yang akan digunakan sebagai dasar untuk perhitungan sebaran radionuklida ke lingkungan.Kata kunci: Inventori, RDE, daya 10 MWt, daya 20 MWt, daya 30 MWt. THE ANALYSIS FOR INVENTORY OF EXPERIMENTAL REACTOR HIGH TEMPERATURE GAS REACTOR TYPE. Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor sourceterm and it will be used as the basis for calculating the distribution of radionuclides into the environment.Keywords: Inventory, RDE, 10 MWt power, 20 MWt power, 30 MWt power.

An Air-Brayton Nuclear-Hydrogen Combined-Cycle Peak- and Base-Load Electric Plant

2007 ◽  
Author(s):  
Charles Forsberg
Keyword(s):  
High Temperature ◽  
Power Output ◽  
Nuclear Reactor ◽  
Peak Load ◽  
Combined Cycle ◽  
Operating Conditions ◽  
Electricity Production ◽  
Spinning Reserve ◽  
High Temperature Reactor ◽  
Base Load

A combined-cycle power plant is proposed that uses heat from a high-temperature nuclear reactor and hydrogen produced by the high-temperature reactor to meet base-load and peak-load electrical demands. For base-load electricity production, air is compressed; flows through a heat exchanger, where it is heated to between 700 and 900°C; and exits through a high-temperature gas turbine to produce electricity. The heat, via an intermediate heat-transport loop, is provided by a high-temperature reactor. The hot exhaust from the Brayton-cycle turbine is then fed to a heat recovery steam generator that provides steam to a steam turbine for added electrical power production. To meet peak electricity demand, after nuclear heating of the compressed air, hydrogen is injected into the combustion chamber, combusts, and heats the air to 1300°C—the operating conditions for a standard natural-gas-fired combined-cycle plant. This process increases the plant efficiency and power output. Hydrogen is produced at night by electrolysis or other methods using energy from the nuclear reactor and is stored until needed. Therefore, the electricity output to the electric grid can vary from zero (i.e., when hydrogen is being produced) to the maximum peak power while the nuclear reactor operates at constant load. Because nuclear heat raises air temperatures above the auto-ignition temperatures of the hydrogen and powers the air compressor, the power output can be varied rapidly (compared with the capabilities of fossil-fired turbines) to meet spinning reserve requirements and stabilize the grid.

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