Session 1: Overviews of High Temperature Engineering Research in Each Country and Organization

Session 1: Overviews of High Temperature Engineering Research in Each Country and Organization

1-1

PRESENT STATUS IN THE NETHERLANDS OF RESEARCH RELEVANT TO HIGH TEMPERATURE IRRADIATION

A.I. van Heek
NRG, P.O. Box 25, 1755 ZG Petten, The Netherlands, email vanheek@nrg-nl.com

R. Conrad
JRC Petten, P.O. Box 2, 1755 ZG Petten, The Netherlands, email conrad@jrc.nl

Because of its favourable design and operational characteristics, and the availability of dedicated experimental equipment, the High Flux Reactor (HFR) in Petten has been extensively used as a test bed for HTR fuel and graphite irradiations for more than 30 years.

The HFR is a multipurpose research reactor, operated at 45 MW for 270 days per year. It is water cooled and water moderated. The variety of neutron field conditions within the tank and outside in the poolside facility make it a very versatile tool for irradiation effects studies. For fuel irradiations, the wide span between high flux density in core positions to low and variable flux density positions in the poolside facility offer the opportunity for different power densities or to adapt the neutron field to the actual burnup state of the specimen under investigation. For rapid radiation damage accumulation in graphite, high flux density central core positions are used with preference.

The earliest HTR fuel testing programmes contributed to the deveopment of the coated particle fuel concept by extended screening tests. Later on, the programmes concentrated on performance testing of reference coated fuel particles and reference fuel elements for the German HTR Module, the HTR-500 and to a lesser extent for the US HTGR concepts. It is shown with representative examples that these fuels have excellent fission product retention capabilities under normal and anticipated off-normal operating conditions.

Extensive irradiation programmes in the HFR Petten have significantly contributed to the database for the design of HTR graphite structures. The programmes not only comprise radiation damage accumulation in the temperature range from 570 to 1570K up to very high fast neutron fluences and its influence on technological properties, but also irradiations under specified load conditions to investigate the irradiation creep behaviour of various graphites in the temperature range 570 to 1170K.

The most recent pebble irradiation experiments aim to investigate the behaviour of SiC coated fuel elements under specific operating conditions of a Modular HTR. The SiC coating layer has been proposed as a corrosion protection layer in the case of air or water ingress Three types of graphite (IG110, V-483 and A3-3) and three coating methods (CVD, slip with Si-infiltration and slip with moistening) were applied, in three experiments of which two are completed and the third will be irradiated in 1999. The first experiment contained seven coated spheres and the second experiment contained five coated spheres, of which three were hollow, and three cylindrical SiC samples. The third experiment that is in preparation is designed for six coated spheres and six cylindrical SiC samples. The irradiation parameters of the three experiments cover typical HTR-Modul operating conditions, ie, fast neutron fluence range from 1.2 - 1.9 10²⁵ m^-2 (E >0.1 MeV) and temperatures ranging from 550 to 800°C.

1-2 ASPECTS OF INHERENT SAFETY OF FUTURE HIGH TEMPERATURE REACTORS

K. Kugeler
Institute for Safety Research and Reactor Technology,
Research Centre Julich, 52425 Julich, Germany

It is imperative, from the viewpoint of future extensive world-wide usage of nuclear technology, to create a "catastrophe-free nuclear technology", i.e. one whereby catastrophe protection plans are no longer necessary. This technology can only be considered successful, if under all possible accident conditions, the radioactive fission products remain entirely within the reactor building and there are no serious consequences for the surroundings of the site: i.e. there are no immediate casualties or detectable late fatalities and no evacuation or relocation of residents is necessary.

There are several reactor designs under development around the world, all with the aim of achieving higher safety standards. In Germany, the changes in the law governing nuclear sites (in 1994) shows commitment to world-wide demands for more stringent safety regulations.

Future nuclear power plants must not only satisfy very high safety requirements, but fulfill a good use of fissile material and allow economical competitiveness for electricity generation, and must offer good opportunity for intermediate and final waste disposal.

On the basis of many years experiences of high temperature reactors a reactor concept can be proposed which cannot suffer any core meltdown even under severe accident conditions and in which the broad know-how can be used world-wide. These reactors can be used in combination with the steam cycle, with the gas turbine cycle and with the combined cycle, too. Furthermore, they are very well suited for co-generation processes, because the thermal power of these reactors will be adapted to the conditions of the energy market. Spherical ceramic fuel elements of HTRs with coated particle can be realized, which remain intact under severe accidents and confine further the fission products in the coated particles up to 1600°C. Decay heat removal is performed self-acting from the core just on the basis of heat conduction, thermal radiation and natural convection of the air. Fuel elements temperature never exceed the value of 1600°C in all accidents.

Inadmissible nuclear transients can never occur, because of strong negative temperature coefficients as well as due to absence of burnup induced excess reactivity in the core, because of the continuous loading of the fuel. However fuel elements could themselves bear severe nuclear transients, if these could occur. Even in these cases the fuel temperature would stay below 1600°C.

Burst-proved reactor pressure vessel take care of the mechanical and chemical stability of the primary system. The ingress of air and steam to the core is governed by the right design.

Such reactors can achieve thermal power up to 500 MW with present realizable dimension of the reactor pressure vessels. A power of 1000 MW can be realized in the future if some progresses in this technology are considered.

Larger power of a plant is realized by modular arrangement of reactors.

1-3 PRESENT STATUS AND FUTURE PROGRAM OF HTTR AND THE INNOVATIVE BASIC RESEARCH ON HIGH TEMPERATURE ENGINEERING

K. Hayashi¹, M.Ishihara¹, S. Ishino², T. Terai3, H. Itoh¹, S. Tagawa⁴,
Y. Katsumura³,M. Yamawaki³, T. Shikama⁵, C. Mori⁶, Y.Miyamoto¹ and T. Tanaka¹
1: Japan Atomic Energy Research Institute, 2: Tokai University, 3: The University of Tokyo,
4: Osaka University, 5: Tohoku University, 6: Aichi Institute of Technology

The High Temperature Engineering Test Reactor (HTTR), which is the first high temperature gas-cooled reactor in Japan, achieved its first criticality on November 10, 1998 at the Oarai Research Establishment of Japan Atomic Energy Research Institute (JAERI). The reactor is now in the stage of commissioning tests at low pow-ers and subsequently at higher powers of 10 and 20 MW, and finally at the full power of 30 MW. Thereafter are followed the performance tests at reactor outlet coolant temperatures of 850 and 950°C, whose successful achievement will officially complete the construction of HTTR in 2001 in an envisaged schedule.

Promotion of the innovative basic researches is one of the main objectives of the construction of HTTR. An overview of the present status and future program of the researches is briefly presented at this meeting. In the field of new materials development, advancement is being made in the preliminary irradiation tests using re-search reactors other than HTTR, on the subjects of 1) neutron-irradiation processing of high-temperature oxide superconductors, 2) neutron transmutation doping (NTD) of silicon carbide (SiC) semiconductor, and 3) a study on the radiation damage mechanism of refractory ceramic composite materials.

In the field of high-temperature radiation chemistry, elaborate works have been proceeded to utilize radiation-induced chemical reactions at high temperatures, in production of SiC fibers from a macromolecule compound of polysilane, as well as in decomposition of heavy oils and plastics. In the fusion technology, preparatory studies have been made to enable in-reactor measurement of property changes of solid tritium breeding materi-als under irradiation.

Finally in the field of high-temperature in-core instrumentation, efforts have been made to develop a heat- and radiation-resistant optical fiber system and other devices which will enable visualization of in-reactor tests, as well as a variety of in-core monitoring of the temperature, dose rate of gamma rays and neutron fluence and other parameters about the irradiation conditions.

Concrete programs of the HTTR irradiation have been devised for development of the new ceramic materials, and now JAERI is preparing the HTTR irradiation in co-operation with co-authors above.

1-4 IDENTIFICATION OF DOMESTIC NEEDS OF MODULAR HTR FOR ELECTRIC AND HEAT PROCESS INDUSTRY IN INDONESIA

Amir Rusli, Bakri Arbie
National Atomic Energy agency for Indonesia
PO BOX 4390, Jakarta 12710, Indonesia

The identification on potential application of modular high temperature gas cooled reactor (HTGR) in Indonesia has been carried out. Identification was done by surveying and analyzing the electric, industrial heat process included captive power for household and industrial complex in 13 region operation areas of PT. PLN-National electric company from west end (Aceh) to east end (Irian Jaya) of 6000 miles away. Surveying was conducted for several parameters such electric and process heat demand in each region included captive power, geological characteristics of region, distance of each region from conventional energy resources and the existing of petrochemical industries or other industries which use high temperature (>500°C). In order to obtain scale of priority in each region, Credit point (1-3) and sensitivity factor (0-1) were applied to each region to get the total significant value. Regions where located in the east cost of Sumatra, noth coast of Java, west-, South- and east coat of Kalimantan, and south coast of Sulawesi, and thousands of small islands beyond them are safe from geological and tectonic point of view. In the preliminary survey and analyzing shown that region III, IV, XII and XIII have a high potential priority followed by region I, II, VI, VIII and IX. The potential of domestic participation in the first and second unit were also investigated in relation with possibility implementing of HTGR project in Indonesia in the future, about 26% and 31% for the total cost for the first and second units respectively. The summary of result as shown in Table 1. Table 1 Summary analysis of potential PLN operation region for adopting HTR

REG.	SITE.	ELCT.	HEAT	CAPT.	DIST.	INDTR.	TOTAL	CLASS.
I	0.2	0.7	0.2	0.3	0.4	0.1	1.90	B
II	0.2	1.05	0.2	0.3	0.2	0.2	2.15	B
III	0.3	1.05	0.3	0.3	0.4	0.3	2.65	A
IV	0.3	1.05	0.3	0.45	0.4	0.2	2.70	A
V	0.3	0.35	0.2	0.3	0.4	0.1	1.65	C
VI	0.3	0.7	0.3	0.45	0.4	0.3	2.45	B
VII	0.1	0.7	0.1	0.15	0.4	0.1	1.55	C
VIII	0.2	0.7	0.2	0.3	0.4	0.1	1.90	B
IX	0.1	0.7	0.2	0.3	0.6	0.1	2.00	B
X	0.2	0.7	0.1	0.15	0.2	0.3	1.65	C
XI	0.1	0.7	0.1	0.15	0.6	0.1	1.75	C
XII	0.2	1.05	0.3	0.45	0.4	0.2	2.60	A
XIII	0.3	1.05	0.3	0.45	0.4	0.2	2.75	A

A (first priority), B (second priority), C (third priority)

1-5 GAS REACTOR AND ASSOCIATED NUCLEAR EXPERIENCE IN THE UK RELEVANT TO HIGH TEMPERATURE REACTOR ENGINEERING

D J Beech, Dr R May
NNC Limited, UK

Until the construction of the Sizewell B Pressurised Water Reactor, all the commercial reactors in the UK were graphite moderated with carbon dioxide as the gas coolant. These first commercial nuclear power stations, built in the late 1950s, were of the Magnox type using natural uranium fuel clad in Magnox ( a magnesium-aluminium alloy). This design was developed over a number of years and a total of eleven stations (including two in Japan and Italy) were built. Work began on the concept of the Advanced Gas Cooled Reactor (AGR) in the early 1960s with the aim of obtaining a higher gas coolant temperature than that of the magnox reactors and thus an increase in the power density of the the core. The AGR programme in the UK culminated in the introduction of the twin AGRs at Heysham and Torness in the late 1980s, with each twin unit station having an output of 1320 MWe. These reactors operated at core outlet temperatures of 650ºC

NNC and its pre-decessor organisations were responsible for the design and build of all of the UK’s commercial Magnox reactors and Advanced Gas-cooled Reactors. The overall experience gained in the UK gas-cooled reactor ( and particularly the AGR) programme is of great value to the development of the HTR. Additionally, NNC was involved at all stages of the DRAGON international OECD project which was one of the principal HTR test facilities until its decommissioning in 1976. NNC’s commercial exploitation of DRAGON technology coupled with the development of its Magnox and AGR experience eventually led to a reference design and full tender submission to the UK Utility for a large-scale HTR design.

Included in the paper will be the experience gained from the design and construction of other national and common reactor designs, including fast reactors, and from their supporting R&D.

This paper will address this experience and its relevance to HTRs which can be summarised under the following headings:

(i) Underlying R&D in the initial AGR component development programme
(ii) Ongoing research work supporting the operating AGR stations and fast reactor development
(iii) The resolution of operational AGR problems
(iv) Work undertaken in support of periodic safety studies and life extension/life-limiting features of the older AGR stations.

Among the individual topics to be addressed will be the following:

High temperature materials
High temperature weld performance including reheat cracking
High temperature oxidation
Graphite oxidation/shrinkage
Tribology

Where applicable, experience-based observations will also be made concerning the behaviour of materials in helium atmospheres. Based on the above, requirements for HTR design development will be discussed.

1-6 A RENEWED INDUSTRIAL INTEREST FOR HTR'S

Michel Lecomte
Framatome, France

Introduction

HTR's have been developed in the 70's and 80's mostly in Germany and the US. Some were successful prototypes (AVR, Peach Bottom 1), others were plagued with technical problems (THTR, Fort Saint Vrain) not uncommon for prototypes and solvable but the adverse political environment of the time cut short their experimental lifetime.

Nevertheless some of the basic qualities of the HTR were demonstrated such as the robust fuel concept, thermodynamic performance, the forgiveness of its behavior during transients... but the costs could not be proven competitive with light water reactors.

Changes in the safety requirements on one hand led to cost increases of LWR and changes in technological developments, on the other hand, led to new HTR concepts that become attractive costwise.

Description of the new HTR concept

The new HTR concept is based on the modular concept the design of which meets the requirement that under no credible circumstance the fuel temperature exceeds 1600°C even without any protection or safeguard system operation. This passive safety requirement can be met by a steel vessel reactor capable of radiating nuclear decay heat in all circumstances and leads to a design with minimal safety systems thereby decreasing the specific cost.

Technological developments performed mostly in the aeronautic industry on gas turbine and compressors similar to jet engines and an efficient gas/gas heat exchanger allow to try to design direct Brayton cycles instead of the classical steam cycle. This change brings two important benefits: a performance increase of 20 % on one hand (48 % efficiency instead of 40 %) and an overall design simplification.

When all the previous factors, safety system simplification, plant design simplification and performance boost are taken into account, the modular HTR designed on a direct cycle becomes economically attractive for a medium size power level of 200 to 300 Mwe.

The Brayton cycle brings other attractive features specific to the HTR. Indeed the waste heat released to the heat sink is, by design, at a medium temperature of 120°C. Such heat is ideally suited for use to feed either district heating systems or multiple effect distillation water desalination plants. Since the heat represents about 50 % of the cost of water, such a design allows for a potential drastic water cost cut.

Medium temperature heat sink release also permits the use of dry cooling towers which brings extra versatility as the location of plants is not linked to water heat sinks.

Finally, the fuel and core design of the HTR are very flexible and allow use of many kinds of fuels from highly enriched uranium to thorium and from pure weapons grade Plutonium to reprocessed light water plutonium.

Conclusion
In conclusion, the safety requirements imposed by the environment, recent technological developments, fuel flexibility and the combination of these features in the gas turbine design all contribute to an attractive economical nuclear plant concept of medium power level. The concept is also well suited to further performance developments.

1-7 REQUIREMENT TO IRRADIATION TESTS OF FUEL ELEMENTS IN THE BASE OF WGPU,
POSSIBILITIES OF RESOURCE TESTS IN RUSSIA

Y. Sukharev
OKB Mechanical Engineering, Russia

(Abstract not available)

1-8

IAEA HIGH TEMPERATURE GAS COOLED REACTOR ACTIVITIES

J. M. Kendall
International Atomic Energy Agency

IAEA activities on high temperature gas cooled reactors are guided by member states, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). A Coordinated Research Project (CRP) on Design and Evaluation of Heat Utilization Systems for the High Temperature Engineering Test Reactor (HTTR) provides supporting technical information for the HTTR undergoing startup testing in Japan. Another CRP on Evaluation of High Temperature Gas Cooled Reactor Performance enhances the international understanding and value of the HTTR and HTR-10 (high temperature reactor under construction in China) programmes. An ongoing activity is directed toward the development and maintenance of an International Database on Irradiated Nuclear Graphite Properties. A Technical Committee Meeting on Safety Related Design and Economic Aspects of High Temperature Gas Cooled Reactors (HTGR) and a Consultancy on Preservation of HTGR Technology and Application were held in November 1998. The meetings resulted in recommendations for additional coordination of international GCR R&D activities, including the dissemination and archiving of information and data related to present and past development work, and development of an information sharing system utilizing recent advancements in electronic communication. A new CRP on conservation and application of HTGR technology, recommended by the meetings, is under development.

1-9 THE RENEWAL OF HTR DEVELOPMENT IN EUROPE

D. Hittner
Framatome, France

I will present in this paper the activities funded by the European Commission in the field of HTR technology that we already had since 1998 and our proposals to the Commission for the next 4 years. I will give a short summary of the main conclusions of the work of our European partnership up to now.

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