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

Abstract

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.

Functional Requirements for the Modeling and Simulation of Advanced (Non-LWR) Reactor Mechanistic Source Term

“…These core materials are including but not limited to a graphite fuel matrix, graphite reflectors, carbon bricks, helium coolant, and control rods and absorber balls containing boron and lithium [9]. Recent studies have been performed focusing on the tritium source in HTGR designs [10], including source term analyses specific to irradiated graphite [11].…”

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

“…HTR-10 is chosen to obtain operation data. The Institute of Nuclear and New Energy Technology (INET), Tsinghua University, has performed several data-collection campaigns on HTR-10, such as source term experimental analysis of irradiated graphite in the core, sampling of the radioactive graphite dust in the primary loop, and R&D of helium sampling loop [19][20][21][22]. INET also has studied many theoretical calculation approaches like the Monte Carlo method and differential equations to predict the radioactivity in HTR-10 [14,17,23,24].…”

Source Term Study on Tritium in HTR-PM: Theoretical Calculations and Experimental Design

“…Thus, residual solid waste will be an important issue for the decommissioning work of HTGRs. Experimental research on H-3 in the irradiated graphite spheres of HTR-10 has been conducted which can supply related information [25]. Current results indicate that most H-3 in the reactor core is generated by the activation reaction of B-10 while most H-3 in the primary coolant comes from activation reactions of He-3 and Li-6.…”