High Temperature Gas-cooled Test Reactor Research Articles

Overview of the 10 MW high temperature gas cooled reactor—test module project

This paper discusses the historical development of the high temperature gas cooled reactor (HTGR) in China. China's development strategy of the HTGR will be explained in this text. The aim, design, construction and commissioning of the 10 MW HTGR—test module (HTR-10) will be explained herein. The engineering experiments, which were developed for the HTR-10, will also be introduced. The experience leading to an accumulation of knowledge during the development of China's HTGR will be summarized in this article.

Structural design and two-phase flow stability test for the steam generator

This paper describes the structural designs and the two-phase flow stability test for the steam generator of the 10 MW high temperature gas-cooled reactor test module. The sectionalized heat transfer surface of the steam generator is made up of modules formed by series-connected single cylindrical coils wound around a displacer. Such a heating surface has some advantages as compared with a heating surface formed by multi concentric coils with a great diameter of winding (Proceeding of the International ANS/ENS Topical Meeting on Thermal Reactor Safety (1986); Progress of HTGR Process Heat Application Study (1994); Heat Transfer-Soviet Res. 22 (1990)). It plays an important part in the design of the integrated steam generator. Since it is a once-through steam generator and operates at a medium pressure (4.0 MPa), two-phase flow instabilities are a difficult problem (Functional Performance of the Helical Coil SG CNSG-IV System (1975)). The design must eliminate two-phase flow instabilities and the system must be tested to show that flow instability does not occur.

The design features of the HTR-10

The 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is a modular pebble bed type reactor. This paper briefly introduces the main design features and safety concept of the HTR-10. The design features of the pebble bed reactor core, the pressure boundary of the primary circuit, the decay heat removal system and the two independent reactor shutdown systems and the barrier of confinement are described in this paper.

Design and tests for the HTR-10 control rod system

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) was brought to criticality in December of 2000. The control rod system discussed in this paper is a part of the HTR-10 to control the reactivity of the core. The stepping motor as the motive power source contributes to simplify the control rod drive mechanism. The sprocket wheel device, an actuator of the control rod, is more suitable for the HTR-10 than other possible designs. A magnetic damper is used to limit the emergency insertion speed of the control rod. Significantly useful information is derived from solution to the lubrication problem under the conditions of high temperature, high radiation, and the inert helium atmosphere. This paper also covers the philosophy behind the design and descriptions of the control rod system and the performance tests.

Design and verification test of the small absorber ball system of the HTR-10

A small absorber ball system is the second shutdown system of the 10 MW High Temperature Gas-cooled Reactor-Test Module. When the control rod system is out of operation, the small absorber ball system is required to emergently shut down the reactor by means of boron carbide balls totally dropping into channels in the side reflector by gravity. Before the reactor re-operates, these small absorber balls must be conveyed from channels in the side reflector back to the ball storage vessels below the head of reactor pressure vessel by a pneumatic conveying process. Detailed design of the small absorber ball system was carried out, and all the components were manufactured. In order to check the performances of the small absorber ball system, a pneumatic conveying experiment and a hot test were carried out in the laboratory, and also a pre-operation of the small absorber ball system was verified on the reactor site at room temperature. Test data of the small absorber ball system show that all the parameters, including time of ball dropping, time of ball conveying back, indication of ball level in the ball storage vessel, gas flowrate, and valve operation states, are acceptable and reasonable. The small absorber ball system has been shown to be capable of performing satisfactorily at operating temperatures of the reactor.

Prediction calculations and experiments for the first criticality of the 10 MW High Temperature Gas-cooled Reactor-Test Module

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is a pebble bed experimental reactor built by the Institute of Nuclear Energy Technology (INET), Tsinghua University. This paper introduces the first critical prediction calculations and the experiments for the HTR-10. The German VSOP neutronics code is used for the prediction calculations of the first loading. The characteristics of pebble-bed high temperature gas-cooled reactors are taken into account, including the double heterogeneity of the fuel element, the buckling feedback of the spectrum calculation, the effect of the mixture of fuel elements and graphite balls, and the correction of the diffusion coefficients in the upper cavity based on transport theory. Also considered are the effects of impurities in the fuel elements, in the graphite balls and in the reflector graphite on the reactivity. The number of fuel elements and graphite balls in the initial core is predicted to provide reference for the first criticality experiment. The critical experiment adopts a method of extrapolating to approach criticality. The first criticality was attained on December 1, 2000. The first criticality experiment shows that the predicted critical number of the fuel elements and graphite balls is in close agreement with the experimental results. Their relative error is less than 1.0%, implying the physical predictions and the results of the criticality experiment are much beyond expectations.

Radioactive Release and Dose Calculations for the HTR-10

A 10-MW high-temperature gas-cooled test reactor, the High-Temperature Reactor-10 (HTR-10), being built at the Institute of Nuclear Energy Technology of Tsinghua University, is a type of modular high-temperature gas-cooled reactor. The design features of the HTR-10 are studied in terms of five important sources of airborne radioactive materials released to the environment. These sources are activation of the air in the reactor cavity, leakage of the primary coolant, release of radioactively contaminated helium from the regeneration of the helium purification systems, release of radioactively contaminated helium from the gas evacuation subsystem of the fuel load and unload systems, and leakage of the vapor from the water/steam loops. On the basis of the HTR-10 design parameters, the amount of radioactivity released to the environment per year is calculated, and the dose to the public is calculated as it relates to the HTR-10 site.

Development of components for the gas-cooled fast breeder reactor program

The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs.