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.

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.

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 Conceptual Core Design Study of the Very High Temperature Gas-Cooled Reactor (VHTR) Upgrading the Core Performance by Using Multihole-Type Fuel

2008 ◽  
Vol 7 (1) ◽  
pp. 32-43 ◽  
Author(s):  
Kazutaka OHASHI ◽  
Tetsuo NISHIHARA ◽  
Kazuhiko KUNITOMI ◽  
Masaaki NAKANO ◽  
Yujiro TAZAWA ◽  
...  
Keyword(s):  
High Temperature ◽  
Design Study ◽  
The Core ◽  
Conceptual Core ◽  
Very High ◽  
Type Fuel ◽  
High Temperature Gas

scholarly journals Management of Radioactive Waste Containing Graphite: Overview of Methods

Energies ◽  
2020 ◽  
Vol 13 (18) ◽  
pp. 4638
Author(s):  
Leon Fuks ◽  
Irena Herdzik-Koniecko ◽  
Katarzyna Kiegiel ◽  
Grazyna Zakrzewska-Koltuniewicz
Keyword(s):  
High Temperature ◽  
Radioactive Waste ◽  
Nuclear Power ◽  
Information Service ◽  
Nuclear Industry ◽  
Current Status ◽  
Triso Fuel ◽  
Irradiated Graphite ◽  
High Temperature Gas

Since the beginning of the nuclear industry, graphite has been widely used as a moderator and reflector of neutrons in nuclear power reactors. Some reactors are relatively old and have already been shut down. As a result, a large amount of irradiated graphite has been generated. Although several thousand papers in the International Nuclear Information Service (INIS) database have discussed the management of radioactive waste containing graphite, knowledge of this problem is not common. The aim of the paper is to present the current status of the methods used in different countries to manage graphite-containing radioactive waste. Attention has been paid to the methods of handling spent TRISO fuel after its discharge from high-temperature gas-cooled reactors (HTGR) reactors.

Study on Emergency Action Levels of the Modular High Temperature Gas-Cooled Reactor

2017 ◽  
Author(s):  
Xiaochuan Zang ◽  
Tao Liu
Keyword(s):  
High Temperature ◽  
Fission Product ◽  
Operating Mode ◽  
Inherent Safety ◽  
Pressurized Water ◽  
Plant Safety ◽  
Water Reactors ◽  
Safety Features ◽  
Action Levels ◽  
High Temperature Gas

The emergency action level (EAL) scheme for a modular high temperature gas-cooled reactor (HTR) plant refers to the generic EAL development guidance for pressurized water reactors (PWR) with HTR modification due to its design issues. Based on reactor’s accidents analysis and consequence assessment, EAL scheme of HTR is established through the steps of category and classification. Four emergency classes are set for HTR consisting of U (Emergency Standby), A (Facilities Emergency), S (Site Area Emergency) and G (General Emergency). The Recognition Category of Initiating Condition (IC) and EAL contains A - Abnormal Rad Levels / Radiological Effluent, F - Fission Product Barrier, H - Hazards and Other Conditions Affecting Plant Safety, S - System Malfunction. The methodology for development of EALs for HTR on Fission Product Barrier and System Malfunction has some differences from PWR’s due to differences on operating mode, inherent safety features and system characteristics.

PKL Tests on Heterogeneous Inherent Boron Dilution Following SB-LOCA: Applicability to Reactor Scale

2010 ◽  
Author(s):  
Klaus Umminger ◽  
Simon Philipp Schollenberger ◽  
Se´bastien Cornille ◽  
Claire Agnoux ◽  
Delphine Quintin ◽  
...  
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Limited Range ◽  
Test Facility ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Physical Phenomena ◽  
Boron Dilution

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.

scholarly journals Study on Air Ingress Mitigation Methods in the Very High Temperature Gas Cooled Reactor (VHTR)

2011 ◽  
Author(s):  
Chang H. Oh ◽  
Eung S. Kim
Keyword(s):  
High Temperature ◽  
Root Cause ◽  
The Core ◽  
Graphite Oxidation ◽  
Simulation Results ◽  
Oxidation Damage ◽  
Air Ingress ◽  
Mitigation Methods ◽  
Very High ◽  
High Temperature Gas

An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR) safety. Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen depending on the accident scenario and the design. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (6) preventing fluid density gradient; (7) preventing fluid temperature gradient; (7) preventing high temperature. Based on the basic concepts listed above, various air-ingress mitigation methods are proposed in this study. Among them, the following one mitigation idea was extensively investigated using computational fluid dynamic codes (CFD) in terms of helium injection in the lower plenum. The main idea of the helium injection method is to replace air in the core and the lower plenum upper part by buoyancy force. This method reduces graphite oxidation damage in the severe locations of the reactor inside. To validate this method, CFD simulations are addressed here. A simple 2-D CFD model was developed based on the GT-MHR 600MWt as a reference design. The simulation results showed that the helium replaces the air flow into the core and significantly reduces the air concentration in the core and bottom reflector potentially protecting oxidation damage. According to the simulation results, even small helium flow was sufficient to remove air in the core, mitigating the air-ingress successfully.

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