In this study several air and water ingress scenarios for the PBMR [1] were simulated by means of the dynamic reactor code TINTE ( TI me-dependent N eutronics and TE mperatures) [2]. The Power Conversion Unit (PCU) and other sub-systems cannot be modelled with the TINTE code and therefore air ingress rates were obtained from Computational Fluid Dynamics (CFD) analysis performed by utilizing the FLUENT code [3]. The use of the TINTE code was previously validated with simulations of the NACOK corrosion experiments [4], [5], [6]. The validations were performed at Forschungszentrum Juelich, however the results are not yet published. The rates of chemical reactions between graphite and gases like O 2 , CO 2 , H 2 O and H 2 are negligible below 400°C. Air and water ingress into the PBMR core at high temperatures can result in corrosion of the PBMR fuel spheres and a possible increase in the fission product release rate. The air ingress scenarios included in this study are; a break in the core outlet pipe at the turbine inlet location, which results in air ingress from the outlet plenum, and a break in a pipe that is connected to the top of the Reactor Pressure Vessel (RPV), which results in air ingress from top of the core. For both transients it is assumed that a Depressurized Loss of Forced Cooling (DLOFC) event takes place prior to the air ingress. The DLOFC leads to high fuel and reflector temperatures that allow higher oxidation rates. The results show that the oxidation of graphite structures in the core is more severe in the case of the outlet pipe break transient. A break in the Core Conditioning System (CCS) heat exchanger circuit during a maintenance mode or following a reactor trip could result in water ingress of up to 1000 kg into the core (the primary system is depressurized at this stage). During the water ingress the CCS continuously cools down the core. Due to the low water ingress rate and lower fuel temperatures, the water ingress transient is not as severe as the air ingress transients.

For direct cycle gas cooled high temperature reactor designs, operating conditions may be limited as a result of excessive maintenance dose rates caused by the Ag-110m source term on the turbine. It is therefore important to accurately predict silver release from fuel during reactor operation. Traditionally diffusion models were used to derive transport parameters from limited irradiation testing of fuel materials and components. Best estimates for all applicable German fuel irradiation tests with defendable uncertainty ranges were never derived. However, diffusion theory and current parameters cannot account for all irradiation and heat-up test results, and for some tests, it appears unacceptably conservative. Other transport mechanisms have been suggested, and alternative calculation models are being considered. In this paper the applicable German irradiation test results are evaluated with a classic diffusion model as well as an alternative model called the Molecular Vapour transport Release (MVR) model. New transport models and parameters for silver in fuel materials are suggested and compared.

The paper deals with the issue of increasing efficiency of nuclear power plants with the modular high-temperature helium reactor (HTGR) and direct gas turbine cycle. It should be noted that only this combination can highlight the advantages of the HTGR, namely the ability to heat helium to about 1000°C, in comparison with other reactor plants for electricity generation. The HTGR has never been used in the direct gas turbine cycle. At present, several designs of such commercial plants are at the stage of experimental validation of main technical features. In Russia, “OKB Mechanical Engineering” together with “General Atomics” (USA) are developing the GT-MHR project with the reactor power of 600 MW, reactor outlet helium temperature of 850 °C, and efficiency of about 45.2%; the South African Republic is developing the PBMR project with the reactor power of 400 MW, reactor outlet helium temperature of 900 °C, and efficiency of about 42%; and Japan is developing the GTHTR-300 project with the reactor power of 600 MW, reactor outlet helium temperature of 850°C, and efficiency of about 45.6%. As it has been proven by technical and economic estimations, one of the most important factors for successful promotion of reactor designs is their net efficiency, which must be not lower than 47%. A significant advantage of a reactor plant with the HTGR and gas-turbine power conversion unit over the steam cycle is considerable simplification of the power unit layout and reduction of the required equipment and systems (no steam generators, no turbine hall including steam lines, condenser, deaerator, etc.), which makes the gas-turbine power conversion unit more compact and less costly in production, operation and maintenance. However, in spite of this advantage, it seems that in the projects currently being developed, the potential of the gas-turbine cycle and high-temperature reactor to more efficiently generate electricity is not fully used. For example, in modern reactor plants with highly recuperative steam cycle with supercritical heat parameters, the net efficiency of electricity generation reaches 50–55%. There are three methods of Brayton cycle carnotization: regeneration, helium cooldown during compression, and heat supply during expansion. These methods can be used both separately and in combination, which gives a total of seven complex heat flow diagrams. Besides, there are ways to increase helium temperature at the reactor inlet and outlet, to reduce hydraulic losses in the helium path, to increase the turbomachine (TM) rotation speed in order to improve the turbine and compressor efficiency, to reduce helium leaks in the circulation path, etc. The analysis of GT-MHR, PBMR and GTHTR-300 development experience allows identification of the main ways of increasing the efficiency by selecting optimal parameters and design solutions for the reactor and power conversion unit. The paper estimates the probability of reaching the maximum electricity generation efficiency in reactor plants with the HTGR and gas turbine cycle with account of the up-to-date development status of major reactor plant components (reactor, vessels, turbocompressor (TC), generator, heat exchange equipment, and structural materials).

The AVR pebble bed reactor (46 MWth) was operated 1967–1988 at coolant outlet temperatures up to 990°C. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published unresolved safety problems of AVR and of pebble bed HTRs. The AVR primary circuit is heavily contaminated with dust bound and mobile metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory. A re-evaluation of the AVR contamination is performed in order to quantify consequences for future HTRs: The AVR contamination was mainly caused by inadmissible high core temperatures, and not — as presumed in the past — by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot be equipped with instruments. The maximum core temperatures were more than 200 K higher than precalculated. Further, azimuthal temperature differences at the active core margin were observed, as unpredictable hot gas currents with temperatures > 1100°C. Despite of remarkable effort these problems are not yet understood. Having the black box character of the AVR core in mind it remains uncertain whether convincing explanations can be found without major experimental R&D. After detection of the inadmissible core temperatures, the AVR hot gas temperatures were strongly reduced for safety reasons. Metallic fission products diffuse in fuel kernel, coatings and graphite and their break through takes place in long term normal operation, if fission product specific temperature limits are exceeded. This is an unresolved weak point of HTRs in contrast to other reactors and is particularly problematic in pebble bed systems with their large dust content. Another disadvantage, responsible for the pronounced AVR contamination, lies in the fact that activity released from fuel elements is distributed in HTRs all over the coolant circuit surfaces and on graphitic dust and accumulates there. Consequences of AVR experience on future reactors are discussed. As long as pebble bed intrinsic reasons for the high AVR temperatures cannot be excluded they have to be conservatively considered in operation and design basis accidents. For an HTR of 400 MWth, 900°C hot gas temperature, modern fuel and 32 fpy the contaminations are expected to approach at least the same order as in AVR end of life. This creates major problems in design basis accidents, for maintenance and dismantling. Application of German dose criteria on advanced pebble bed reactors leads to the conclusion that a pebble bed HTR needs a gas tight containment even if inadmissible high temperatures as observed in AVR are not considered. However, a gas tight containment does not diminish the consequences of the primary circuit contamination on maintenance and dismantling. Thus complementary measures are discussed. A reduction of demands on future reactors (hot gas temperatures, fuel burn-up) is one option; another one is an elaborate R&D program for solution of unresolved problems related to operation and design basis accidents. These problems are listed in the paper.

This paper discusses the reliability and integrity management (RIM) strategies that have been applied in the design of the PBMR passive metallic components for the helium pressure boundary (HPB) to meet reliability targets and to evaluate what combination of strategies are needed to meet the targets. The strategies considered include deterministic design strategies to reduce or eliminate the potential for specific damage mechanisms, use of an on-line leak monitoring system and associated design provisions that provide a high degree of leak detection reliability, and periodic non-destructive examinations combined with repair and replacement strategies to reduce the probability that degradation would lead to pipe ruptures. The PBMR RIM program for passive metallic piping components uses a leak-before-break philosophy. A Markov model developed for use in LWR risk-informed inservice inspection evaluations was applied to investigate the impact of alternative RIM strategies and plant age assumptions on the pipe rupture frequencies as a function of rupture size. Some key results of this investigation are presented in this paper.

The Next Generation Nuclear Plant (NGNP) Project is a US-based initiative led by Idaho National Laboratories to demonstrate the viability of using High Temperature Gas-Cooled Reactor (HTGR) technology for the production of high temperature steam and/or heat for applications such as heavy oil recovery, process steam/cogeneration and hydrogen production. A key part of the NGNP Project is the development of a Component Test Facility (CTF) that will support the development of high temperature gas thermal-hydraulic technologies as applied in heat transport and heat transfer applications in HTGRs. These applications include, but are not limited to, primary and secondary coolants, direct cycle power conversion, co-generation, intermediate, secondary and tertiary heat transfer, demonstration of processes requiring high temperatures as well as testing of NGNP specific control, maintenance and inspection philosophies and techniques. The feasibility of the envisioned CTF as a development and testing platform for components and systems in support of the NGNP was evaluated. For components and systems to be integrated into the NGNP full scale or at least representative size tests need to be conducted at NGNP representative conditions, with regards to pressure, flow rate and temperature. Typical components to be tested in the CTF include heat exchangers, steam generators, circulators, valves and gas piping. The Design Data Needs (DDNs), Technology Readiness Levels (TRLs) as well as Design Readiness Levels (DRLs) prepared in the pre-conceptual design of the NGNP Project and the NGNP lifecycle requirements were used as inputs to establish the CTF Functional and Operating Requirements (F&ORs). The existing South African PBMR test facilities were evaluated to determine their current applicability or possible modifications to meet the F&ORs of the CTF. Three concepts were proposed and initial energy balances and layouts were developed. This paper will present the results of this CTF study and the ongoing efforts to establish the CTF.