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Highlights

30/01/15 The fourth ARCHER Newsletter was released on 30 January 2015. Read online here

21/01/15 The second ARCHER EUROCOURSE, hosted by NRG, was held in Petten from 19-20 January 2015. Click here for more info

26/11/14 The third ARCHER Newsletter was released on 26 November 2014. Read online here

28/10/14 The ARCHER final meeting was held on 21-22 Jan 2015 at NRG in Petten (NL).

27/10/14 The High Temperature Reactor (HTR) Conference was held from 27-31 October 2014 in Weihai, Shandong Province, China. Click here for more info

29/03/14 The second Newsletter was released on 28 March 2014. Read online here


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Supported by

European Commission

Furthering key safety aspects of an HTR

Subproject 2 of the ARCHER Project dealt with key safety aspects of the primary circuit and the secondary circuit of an HTR. It built on results of the RAPHAEL Project as documented in the Recommendations of the RAPHAEL Safety Advisory Group, on experience from past and recent licensing activities for HTGR and on public discussion of HTGR safety aspects. Accordingly, the subproject was subdivided in 5 work packages on air ingress, water or steam ingress, the source term analysis chain, thermal safety issues, and the integrity concept of the primary (helium) pressure boundary. In these work packages, experimental tasks to broaden the data base for future code validation or safety assessments as well as computational and other safety analyses were performed. In the following, the major tasks and achievements of the work packages are briefly described.

Regarding air ingress, experiments have been performed in the existing NACOK facility at FZJ. Since air ingress in general has been investigated in the past in several experiments, the tests under this work package focused on a better understanding of the secondary Bordouard reaction in which carbon dioxide and carbon react at high temperatures to carbon monoxide. This reaction is safety relevant because it attacks the surface of the fuel elements and because carbon monoxide may form burnable or explosive gas mixtures with the air inside the reactor building. The tests have been completed successfully for two set-ups, a stack of block type graphitic elements and a pebble bed. Pre- and post-calculations have been performed with a dedicated computer code.

Water or steam ingress into an HTR reactor core is safety relevant for three reasons: steam will increase the reactivity of HTR cores which are usually under-moderated, steam can cause corrosion of graphite, albeit less than air, and wash out fission products from defect coated particles which are present in the core in low fractions (<10-4). The gaseous reaction products, especially hydogen, may form flammable or explosive mixtures with the air of the reactor building if released through the primary pressure relief valves. Since in the last decades the focus of HTR development was on systems with gas turbines in direct cycle and on process heat application with intermediate heat exchangers, and only in the last years steam production in a secondary cycle gained attention again, this work package attempted to update knowledge and data base with regard to water / steam ingress into an HTR core. The flammability limits of water ingress exhaust gas mixtures in air were experimentally determined and flammability diagrams produced.



Existing computer models were refined and validated to treat simultaneously the nucleonic, thermal hydraulic and corrosive effects of steam ingress into the core. For the design basis event: instantaneous rupture of one steam generator tube, it could be shown that former assumptions of 600 kg of steam entering the core and reacting there were conservative. The detailed calculations performed now show that less than half of this value would need to be considered.

During the operation of a nuclear reactor, radioactive fission products and activation products are produced. Their release to the environment has to be avoided or limited to acceptably low quantities. Regulatory limits are set in countries using nuclear power and/or are also proposed by international organisations. In order to demonstrate compliance with these limits, the entire way of the radioactive material from their place of origin to the environment has to be analysed.This is the source term analysis chain mentioned above. For the HTR it turned out in the recent years, that the understanding of the behaviour of graphitic dust which is produced in the reactor needs to be improved. This dust is loaded with radioactive fission products and activation products and would, if released to the environment, contribute to the radioactive source term. For these reasons, the majority of tasks in the third work package was concerned with different aspects of the behaviour of dust along the source term analysis chain.

Firstly, in a small experimental device called pebble mill, dust was produced by friction between typical graphitic material grades which were used in HTR or are likely to be used in the future. Size and shape of the generated dust particles were analysed as relevant input to computer simulations. Two other tasks investigated the deposition and remobilisation of dust in simple geometries, again in order to provide data for future validation of computer codes. In a third experimental task the deposition and remobilisation of dust in a pebble itself were investigated. Radioactively marked dust was deposited by an airflow in a pebble bed and subsequently remobilised by increasing the air flow. The growth and decrease of dust layers on the pebbles was monitored by Positron Emission Tomography (PET). It turned out that above a remobilisation flow threshold velocity of about twice the deposition velocity, about 40% of the deposited dust are soon remobilised while the rest adheres to the pebbles.

A good understanding of the deposition behaviour of dust in the primary circuit and its potential remobilisation in the course of breaks of the primary pressure boundary is important, since only that fraction of the dust that is remobilised again can contibute to the source term.

Futhermore the behaviour of the dust in the reactor building, once it has been released from the primary pressure boundary, is important. For this part of the analysis chain, the existing computer code COCOSYS for the analysis of LWR containments has been enhanced to consider dust in the primary gas entering the reactor building. Calculations for two different scenarios with low and high gas flow rates into the reactor building showed that roughly half of the dust entering the building would be deposited there and would thus not contribute to the source term to the environment.

Another task in this work package dealt with the numerical simulation of dust behaviour in pebble beds in computed fluid dynamics (CFD) calculations. A simplified, but with modern computers manageable approach, the Reynolds Averaged Navier Stokes (RANS) approach was validated against available experiments and against the very time consuming Direct Numerical Simulation (DNS) approach, and was successfully applied to ordered and randomly packed pebble beds.

Apart from the questions concerning dust behaviour, Tritium was treated in the last task of this work package. Tritium is produced in the reactor core and the graphite reflectors and migrates rather easily through metallic walls. Its concentration in the secondary steam circuit has to be limited in order to limit doses to plant personnel (closed steam cycle) or to limit doses to the public (open steam cycle for process heat applications). It turned out that in both cases a viable Tritium control strategy would not be problematic to achieve.

A numerical analysis of fission product releases from the fuel elements of the Chinese HTR-10 has also been performed taking into account available information about its operating history.

The 4th Work Package dealt with specific thermal safety issues of HTR. One of them is the fact that fuel element temperatures cannot be measured directly and continuously in a pebble bed. In order to verify the maximum operating fuel temperatures in a pebble bed HTR, the AVR in Germany had been equipped twice in its lifetime with instrumented spheres, in which melt wires melting at a range of specified temperatures were embedded. The experiments showed that the AVR had been run at significantly higher maximum core temperatures than calculated. Different causes have been discussed in the last years, ranging from 3D-effects caused by the reflector noses (an AVR specific design feature) to generally high flow bypasses to the core. In preparation of new melt wire tests in a current pebble bed HTR (HTR-10 in China was envisaged, but could not be used due to occupation of INET with the design of the HTR-PM), a specification for new melt wire spheres was produced together with proposals for conduct of the experiment in the reactor. In two other tasks, variations in the porosity (or packing fraction (PF)) of the pebble have been investigated experimentally and numerically as another potential candidate for local overtemperatures in a pebble bed core. It turned out that such variations exist, but their spatial dimension is rather small, so that there are no significant effects on the maximum fuel temperatures, neither during operation nor under accident conditions.



The last Work Package dealt with the integrity of the primary pressure boundary. In order to exclude large air ingress leading to core damage, large or multiple breaks of the pressure boundary must be excluded. Such a break exclusion has been achieved in European LWR, and the assessment in this work package concluded that it would also be applicable with minor additional effort to HTR primary pressure boundaries, provided the same material is used (as was the case with European Modular HTR concepts and is the case with the Chinese HTR-PM currently under construction). Additional efforts would mainly be restricted to broaden the data base regarding corrrosion in a Helium atmosphere with low impurities. Existing data indicate that this should be no problem in the temperature regime of small modular HTR.

When the ARCHER project was defined, proposals were made for larger modular HTR to employ a primary pressure boundary made of modified 9% chromium (P91) steel and operate it at temperatures in the creep range. Invoking break exclusion for such a concept would require significantly higher R&D efforts.