Arbeitsgemeinschaft Versuchsreaktor Research Articles

Effects of %FIMA on Storage-Safety Parameters of Spent Fuel from Experimental Pebble-Bed Reactor

The back end of the utilization of nuclear technology is safety and management of spent fuel, which is a key element contributing to the success of the nuclear power plant program. Indonesia’s National Nuclear Energy Agency resolved to establish an experimental power reactor, called RDE, as a nuclear power plant demo. The fuel of this reactor is similar to that of German’s experimental pebble-bed reactor (PBR), Arbeitsgemeinschaft Versuchsreaktor(AVR). In this study, the spent fuel of AVR was studied to obtain the safety parameter data for storage of RDE spent fuel by varying the fission in the initial metallic atoms (%FIMA). These parameters that must be studied include the radioactivity, decay heat, proliferation threats of both 239Pu and 235U, and the presence of 137Cs, a dangerous fission product that can escape from damaged spent fuels. The calculation was conducted by ORIGEN 2.1. The result of the study demonstrates a higher %FIMA indicates a higher safety level that is required since the activity and decay heat of the spent fuel will increase and, as will be the total amounts of 239Pu and 137Cs. However, the 235U amount will decrease. For a 100 years storage of spent fuel, the optimum %FIMA is 8.2 with a canister capacity of 1,900 pebbles. Further, the activity and decay heat of the spent nuclear fuel are 2.013 × 1013 Bq and 6.065 W, respectively. The activities of 239Pu, 137Cs, and 235U are 5.187 ×1011, 7.100 × 1012, and 7.339 × 107 Bq, respectively.

The Operator Training Simulator System for the Pebble Bed Modular Reactor (PBMR) Plant

The Pebble Bed Modular Reactor (PBMR) is a First of a Kind Engineering with respect to the over 200 new innovations used in the design. The PBMR technical design is an inherited modified design from an earlier design such as the German 15 MWe AVR (Arbeitsgemeinschaft Versuchs Reaktor) and the THTR-300 MWe Thorium High Temperature Reactor (THTR), which ran in Germany as a test and research facility for 20 years. This paper discusses the Operator Training Simulator System for the PBMR Demonstration Power Plant. The Operator Training Simulator System will be used for operator training and licensing of plant operators. Included in the discussion is an overview of the major elements of the Operator Training Simulator System, including some of the main functional areas.

Technology and Safety Aspects of the German High-Temperature Gas-Cooled Reactor

In the Federal Republic of Germany, hightemperature gas-cooled reactor (HTGR) operating experience with the experimental Arbeitsgemeinschaft Versuchsreaktor (AVR) and the thorium high-temperature r...

Contributions towards the development of a packaging concept for the final disposal of spent HTGR pebble bed fuel

In accordance with decisions of the Government of the German Federal Republic on the treatment of spent fuel from nuclear reactors, the spent spherical fuel elements from the German pebble bed High Temperature Gascooled Reactors (HTGRs). AVR (Arbeitsgemeinschaft Versuchs Reaktor) and THTR (Thorium HochTemperatur Reaktor) shall be directly disposed of in the national repository. Owing to their radiological and thermal properties, the spent HTGR fuel may be treated similar to heat generating medium active waste for which the disposal concept is the emplacement in deep boreholes of a salt mine. Packaging concepts for the final disposal of spent HTGR fuel shall follow the general design and the radiological requirements for the borehole emplacement technique, but credit may also be taken of advantageous properties of the HTGR fuel elements. Especially their built-in barriers against fission product release, designed for stringent operational, upset or even accident conditions in service, can help in the development of a safe and economic packaging and disposal concept for spent HTGR fuel. Evaluation of the mechanical behaviour of HTGR spent fuel elements in experimental testing under near repository conditions has shown that the standard HTGR fuel particle can generally survive the crushing of the elements under rock pressure, but, for safety reasons, a backfill matrix such as quartz sand or cement should be introduced in the case of thin walled disposal canisters to prevent element breakage as a whole at rock pressure.

The TINTE Modular Code System for Computational Simulation of Transient Processes in the Primary Circuit of a Pebble-Bed High-Temperature Gas-Cooled Reactor

The TINTE code deals with the nuclear and thermal transient behavior of a high-temperature reactor, taking into consideration the mutual feedback effects in two-dimensional r-z geometry, including the following: 1. time-dependent neutron flux calculation2. time-dependent heat source distribution (local and nonlocal fractions)3. time-dependent heat transport from the fuel to the fuel element surface4. time-dependent global temperature distribution5. gas flow distribution for a given mass flow or given pressure difference or even under natural circulation conditions6. convection and its feedback to the circulation. The TINTE code has been tested against transient experiments performed at the Arbeitsgemein-schaft Versuchsreaktor. It has also been used successfully for other reactor investigations and thermofluid problems including design studies for a small heating reactor of the peu à peu loading type.

Planning and Performance of Dynamic Experiments with the AVR Power Plant

In collaboration with Kernforschungsanlage Julich, Federal Republic of Germany, and other companies, dynamic experiments have been carried out with the Arbeitsgemeinschaft Versuchsreaktor (AVR) to test advanced dynamic computer models with the goal of using low-enriched uranium (LEU) fuel in future high-temperature gas-cooled reactors. Since LEU fuel has been used for the AVR since 1982, both experimental and theoretical behavior has been studied during the changeover from highly enriched uranium to LEU. The experiments comprise fast power transients that are initiated by either a fast control rod movement or a fast change of coolant flow. The neutron flux and other important parameters are registered in suitable time expansion.

Neutron Physical Investigations on the Shutdown Effect of Small Boronated Absorbing Spheres for Pebble-Bed High-Temperature Gas-Cooled Reactors

An emergency shutdown system for high-temperature gas-cooled pebble-bed reactors is proposed in addition to the common absorber rod shutdown system. This system is based on the strongly absorbing effect of small boronated graphite spheres (called KLAK), which trickle in case of emergency by gravity from the top reflector into the reactor core. The inner reflector of the Siemens-Argonaut reactor was substituted by an assembly of spherical Arbeitsgemeinschaft Versuchsreaktor fuel elements, and the shutdown effect was examined by installing well-defined KLAK nests inside this assembly. The purpose was to develop and prove a calculational procedure for determining criticality values for assemblies of large fuel spheres and small absorbing spheres.

Analysis of Reactivity and Temperature Transient Experiments at the AVR High-Temperature Reactor

In the Arbeitsgemeinschaft Versuchsreaktor high-temperature reactor low-enriched fuel elements have been loaded for several years, replacing about half of the initial highly enriched uranium-thorium fuel. This changeover is accompanied by an elaborate experimental program wherein changes in the neutronic and thermohydraulics behavior of the system are monitored. Within this program reactor transients are induced by means of either circulator speed reduction or control rod movement. The resulting transients in neutron flux and temperatures are recorded and used as a basis for detailed theoretical analysis. Working groups at several institutions have carried out transient calculations using different dynamic codes. The results are presented and differences are discussed. In general, the experimental values were reproduced in a very satisfactory way.

Nuclear Energy for Coal and Steel

Besides the production of electricity, the high-temperature reactor (HTR) offers the potential for producing secondary energy carriers for the fuel and heat market. Therefore, the HTR can make a considerable contribution to solving future problems in the energy supply of the Federal Republic of Germany, as well as of the whole world. On the basis of experience with the power plants Arbeitsgemeinschaft Versuchsreaktor, Fort St. Vrain, and THTR-300, new concepts of reactors have been proposed: the medium-sized reactor HTR 500 and the modular HTR concept. The high-temperature heat application is directed toward the refinement of fossil fuels, the long-distance energy system, and other applications, such as process steam for the chemical industry, enhanced oil recovery, and energy for steel production. The research and development program in the Prototype Plant Nuclear Process Heat and Nuclear Long-Distance Energy projects has shown very promising results. These results show that nuclear process heat is technically feasible and that it is possible to reach a commercial application in the next few decades.

Status of Qualification of High-Temperature Reactor Fuel Element Spheres

To cover a large number of possible high-temperature reactor applications in the Federal Republic of Germany, the fuel element is designed to be uniform and acceptable for conditions set by different approaches for various plants following the thorium high-temperature reactor (THTR). To minimize fission product release, silicon carbide coated fuel (Triso) is required for all types of fuel cycles, and limits are specified for both fabrication and in-service coating defects. Fabrication processes using precise classification of coated and overcoated particles by tabling techniques ensure satisfaction of specified limits on fabrication induced coating defects. A comprehensive irraditation test program was performed for both high-enriched uranium/thorium fuel elements used in the THTR and Arbeitsgemeinschaft Versuchs-Reaktor cycle and the low-enriched uranium fuel elements to be used for follow-up projects. The results demonstrate good performance of all types of fuel.

Preliminary operating experiences with the AVR at an average hot-gas temperature of 950°C

The two-loop system with a high temperature reactor, which is operated by the Arbeitsgemeinschaft Versuchsreaktor (AVR) GmbH and which was built by the BBC/Krupp consortium (today HRB), has been in operation for more than seven years. In that time more than 635 × 10 6 kWhr have been produced and more than 6.5 × 10 5 spherical fuel elements have been circulated under operation. The fully integrated design, and above all the ceramic gas duct, permit very high gas temperatures although no high alloyed, heat resistant steels were used in the reactor. In February 1974 the average hot-gas temperature at the outlet of the core could thus be increased from its original design value of 850°C to 950°C. Peak temperatures of above 500°C are thereby confined to a small region between the middle of the core and the beginning of the steam generator. Carbon protects the steel structures against high temperatures. Unplanned interruptions and reductions of operation due to the increase of the hot-gas temperature have not occurred so far. Some thermocouples in the hot-gas region failed. All other components functioned satisfactorily and, one year after the increase in the hot-gas temperature, there are no misgivings as to their future functioning. These satis-factory but short operating experiences at 950°C will have to be supplemented in the next few years by experiences over a longer period.